WorldWideScience

Sample records for air cooled reactors

  1. Steady Thermal Field Simulation of Forced Air-cooled Column-type Air-core Reactor

    Institute of Scientific and Technical Information of China (English)

    DENG Qiu; LI Zhenbiao; YIN Xiaogen; YUAN Zhao

    2013-01-01

    Modeling the steady thermal field of the column-type air-core reactor,and further analyzing its distribution regularity,will help optimizing reactor design as well as improving its quality.The operation mechanism and inner insulation structure of a novel current limiting column-type air-core reactor is introduced in this paper.The finite element model of five encapsulation forced air-cooled column type air-core reactor is constructed using Fluent.Most importantly,this paper present a new method that,the steady thermal field of reactor working under forced air-cooled condition is simulated without arbitrarily defining the convection heat transfer coefficient for the initial condition; The result of the thermal field distribution shows that,the maximum steady temperature rise of forced air-cooled columntype air-core reactor happens approximately 5% to its top.The law of temperature distribution indicates:In the 1/3part of the reactor to its bottom,the temperature will rise rapidly to the increasing of height,yet the gradient rate is gradually decreasing; In the 5 % part of the reactor to its top,the temperature will drop rapidly to the increasing of height; In the part between,the temperature will rise slowly to the increasing of height.The conclusion draws that more thermal withstand capacity should be considered at the 5 % part of the reactor to its top to achieve optimal design solution.

  2. Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

    Energy Technology Data Exchange (ETDEWEB)

    Sulaiman, S. A., E-mail: shamsulamri@tamu.edu; Dominguez-Ontiveros, E. E., E-mail: elvisdom@tamu.edu; Alhashimi, T., E-mail: jbudd123@tamu.edu; Budd, J. L., E-mail: dubaiboy@tamu.edu; Matos, M. D., E-mail: mailgoeshere@gmail.com; Hassan, Y. A., E-mail: yhasssan@tamu.edu [Department of Nuclear Engineering, Texas A and M University, College Station, TX, 77843-3133 (United States)

    2015-04-29

    The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A and M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

  3. Design considerations and experimental observations for the TAMU air-cooled reactor cavity cooling system for the VHTR

    Science.gov (United States)

    Sulaiman, S. A.; Dominguez-Ontiveros, E. E.; Alhashimi, T.; Budd, J. L.; Matos, M. D.; Hassan, Y. A.

    2015-04-01

    The Reactor Cavity Cooling System (RCCS) is a promising passive decay heat removal system for the Very High Temperature Reactor (VHTR) to ensure reliability of the transfer of the core residual and decay heat to the environment under all off-normal circumstances. A small scale experimental test facility was constructed at Texas A&M University (TAMU) to study pertinent multifaceted thermal hydraulic phenomena in the air-cooled reactor cavity cooling system (RCCS) design based on the General Atomics (GA) concept for the Modular High Temperature Gas-Cooled Reactor (MHTGR). The TAMU Air-Cooled Experimental Test Facility is ⅛ scale from the proposed GA-MHTGR design. Groundwork for experimental investigations focusing into the complex turbulence mixing flow behavior inside the upper plenum is currently underway. The following paper illustrates some of the chief design considerations used in construction of the experimental test facility, complete with an outline of the planned instrumentation and data acquisition methods. Computational Fluid Dynamics (CFD) simulations were carried out to furnish some insights on the overall behavior of the air flow in the system. CFD simulations assisted the placement of the flow measurement sensors location. Preliminary experimental observations of experiments at 120oC inlet temperature suggested the presence of flow reversal for cases involving single active riser at both 5 m/s and 2.25 m/s, respectively and four active risers at 2.25 m/s. Flow reversal may lead to thermal stratification inside the upper plenum by means of steady state temperature measurements. A Particle Image Velocimetry (PIV) experiment was carried out to furnish some insight on flow patterns and directions.

  4. Gas cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1972-06-01

    Although most of the development work on fast breeder reactors has been devoted to the use of liquid metal cooling, interest has been expressed for a number of years in alternative breeder concepts using other coolants. One of a number of concepts in which interest has been retained is the Gas-Cooled Fast Reactor (GCFR). As presently envisioned, it would operate on the uranium-plutonium mixed oxide fuel cycle, similar to that used in the Liquid Metal Fast Breeder Reactor (LMFBR), and would use helium gas as the coolant.

  5. Thermal-Hydraulic Analysis of an Experimental Reactor Cavity Cooling System with Air. Part I: Experiments; Part II: Separate Effects Tests and Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Corradin, Michael [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Anderson, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Muci, M. [Univ. of Wisconsin, Madison, WI (United States). Dept. of Engineering Physics; Hassan, Yassin [Texas A & M Univ., College Station, TX (United States); Dominguez, A. [Texas A & M Univ., College Station, TX (United States); Tokuhiro, Akira [Univ. of Idaho, Moscow, ID (United States); Hamman, K. [Univ. of Idaho, Moscow, ID (United States)

    2014-10-15

    This experimental study investigates the thermal hydraulic behavior and the heat removal performance for a scaled Reactor Cavity Cooling System (RCCS) with air. A quarter-scale RCCS facility was designed and built based on a full-scale General Atomics (GA) RCCS design concept for the Modular High Temperature Gas Reactor (MHTGR). The GA RCCS is a passive cooling system that draws in air to use as the cooling fluid to remove heat radiated from the reactor pressure vessel to the air-cooled riser tubes and discharged the heated air into the atmosphere. Scaling laws were used to preserve key aspects and to maintain similarity. The scaled air RCCS facility at UW-Madison is a quarter-scale reduced length experiment housing six riser ducts that represent a 9.5° sector slice of the full-scale GA air RCCS concept. Radiant heaters were used to simulate the heat radiation from the reactor pressure vessel. The maximum power that can be achieved with the radiant heaters is 40 kW with a peak heat flux of 25 kW per meter squared. The quarter-scale RCCS was run under different heat loading cases and operated successfully. Instabilities were observed in some experiments in which one of the two exhaust ducts experienced a flow reversal for a period of time. The data and analysis presented show that the RCCS has promising potential to be a decay heat removal system during an accident scenario.

  6. MEANS FOR COOLING REACTORS

    Science.gov (United States)

    Wheeler, J.A.

    1957-11-01

    A design of a reactor is presented in which the fuel elements may be immersed in a liquid coolant when desired without the necessity of removing them from the reactor structure. The fuel elements, containing the fissionable material are in plate form and are disposed within spaced slots in a moderator material, such as graphite to form the core. Adjacent the core is a tank containing the liquid coolant. The fuel elements are mounted in spaced relationship on a rotatable shaft which is located between the core and the tank so that by rotation of the shaft the fuel elements may be either inserted in the slots in the core to sustain a chain reaction or immersed in the coolant.

  7. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

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

  8. Small Liquid Metal Cooled Reactor Safety Study

    Energy Technology Data Exchange (ETDEWEB)

    Minato, A; Ueda, N; Wade, D; Greenspan, E; Brown, N

    2005-11-02

    The Small Liquid Metal Cooled Reactor Safety Study documents results from activities conducted under Small Liquid Metal Fast Reactor Coordination Program (SLMFR-CP) Agreement, January 2004, between the Central Research Institute of the Electric Power Industry (CRIEPI) of Japan and the Lawrence Livermore National Laboratory (LLNL)[1]. Evaluations were completed on topics that are important to the safety of small sodium cooled and lead alloy cooled reactors. CRIEPI investigated approaches for evaluating postulated severe accidents using the CANIS computer code. The methods being developed are improvements on codes such as SAS 4A used in the US to analyze sodium cooled reactors and they depend on calibration using safety testing of metal fuel that has been completed in the TREAT facility. The 4S and the small lead cooled reactors in the US are being designed to preclude core disruption from all mechanistic scenarios, including selected unprotected transients. However, postulated core disruption is being evaluated to support the risk analysis. Argonne National Laboratory and the University of California Berkeley also supported LLNL with evaluation of cores with small positive void worth and core designs that would limit void worth. Assessments were also completed for lead cooled reactors in the following areas: (1) continuing operations with cladding failure, (2) large bubbles passing through the core and (3) recommendations concerning reflector control. The design approach used in the US emphasizes reducing the reactivity in the control mechanisms with core designs that have essentially no, or a very small, reactivity change over the core life. This leads to some positive void worth in the core that is not considered to be safety problem because of the inability to identify scenarios that would lead to voiding of lead. It is also believed that the void worth will not dominate the severe accident analysis. The approach used by 4S requires negative void worth throughout

  9. A gas-cooled reactor surface power system

    Science.gov (United States)

    Lipinski, Ronald J.; Wright, Steven A.; Lenard, Roger X.; Harms, Gary A.

    1999-01-01

    A human outpost on Mars requires plentiful power to assure survival of the astronauts. Anywhere from 50 to 500 kW of electric power (kWe) will be needed, depending on the number of astronauts, level of scientific activity, and life-cycle closure desired. This paper describes a 250-kWe power system based on a gas-cooled nuclear reactor with a recuperated closed Brayton cycle conversion system. The design draws upon the extensive data and engineering experience developed under the various high-temperature gas cooled reactor programs and under the SP-100 program. The reactor core is similar in power and size to the research reactors found on numerous university campuses. The fuel is uranium nitride clad in Nb1%Zr, which has been extensively tested under the SP-100 program. The fuel rods are arranged in a hexagonal array within a BeO block. The BeO softens the spectrum, allowing better use of the fuel and stabilizing the geometry against deformation during impact or other loadings. The system has a negative temperature feedback coefficient so that the power level will automatically follow a variable load without the need for continuous adjustment of control elements. Waste heat is removed by an air-cooled heat exchanger using cold Martian air. The amount of radioactivity in the reactor at launch is very small (less than a Curie, and about equal to a truckload of uranium ore). The system will need to be engineered so that criticality can not occur for any launch accident. This system is also adaptable for electric propulsion or life-support during transit to and from Mars.

  10. System and method for air temperature control in an oxygen transport membrane based reactor

    Science.gov (United States)

    Kelly, Sean M

    2016-09-27

    A system and method for air temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

  11. Liquid Metal Cooled Reactor for Space Power

    Science.gov (United States)

    Weitzberg, Abraham

    2003-01-01

    The conceptual design is for a liquid metal (LM) cooled nuclear reactor that would provide heat to a closed Brayton cycle (CBC) power conversion subsystem to provide electricity for electric propulsion thrusters and spacecraft power. The baseline power level is 100 kWe to the user. For long term power generation, UN pin fuel with Nb1Zr alloy cladding was selected. As part of the SP-100 Program this fuel demonstrated lifetime with greater than six atom percent burnup, at temperatures in the range of 1400-1500 K. The CBC subsystem was selected because of the performance and lifetime database from commercial and aircraft applications and from prior NASA and DOE space programs. The high efficiency of the CBC also allows the reactor to operate at relatively low power levels over its 15-year life, minimizing the long-term power density and temperature of the fuel. The scope of this paper is limited to only the nuclear components that provide heated helium-xenon gas to the CBC subsystem. The principal challenge for the LM reactor concept was to design the reactor core, shield and primary heat transport subsystems to meet mission requirements in a low mass configuration. The LM concept design approach was to assemble components from prior programs and, with minimum change, determine if the system met the objective of the study. All of the components are based on technologies having substantial data bases. Nuclear, thermalhydraulic, stress, and shielding analyses were performed using available computer codes. Neutronics issues included maintaining adequate operating and shutdown reactivities, even under accident conditions. Thermalhydraulic and stress analyses calculated fuel and material temperatures, coolant flows and temperatures, and thermal stresses in the fuel pins, components and structures. Using conservative design assumptions and practices, consistent with the detailed design work performed during the SP-100 Program, the mass of the reactor, shield, primary heat

  12. Air cooled absorption chillers for solar cooling applications

    Science.gov (United States)

    Biermann, W. J.; Reimann, R. C.

    1982-03-01

    The chemical composition of a 'best' absorption refrigerant system is identified, and those properties of the system necessary to design hot water operated, air cooled chilling equipment are determined. Air cooled chillers from single family residential sizes into the commercial rooftop size range are designed and operated.

  13. Electrochemistry of Water-Cooled Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Macdonald, Dgiby; Urquidi-Macdonald, Mirna; Pitt, Jonathan

    2006-08-08

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

  14. Air-cooled, hydrogen-air fuel cell

    Science.gov (United States)

    Shelekhin, Alexander B. (Inventor); Bushnell, Calvin L. (Inventor); Pien, Michael S. (Inventor)

    1999-01-01

    An air-cooled, hydrogen-air solid polymer electrolyte (SPE) fuel cell with a membrane electrode assembly operatively associated with a fluid flow plate having at least one plate cooling channel extending through the plate and at least one air distribution hole extending from a surface of the cathode flow field into the plate cooling channel.

  15. Application of Hastelloy X in Gas-Cooled Reactor Systems

    DEFF Research Database (Denmark)

    Brinkman, C. R.; Rittenhouse, P. L.; Corwin, W.R.

    1976-01-01

    Hastelloy X, an Ni--Cr--Fe--Mo alloy, may be an important structural alloy for components of gas-cooled reactor systems. Expected applications of this alloy in the High-Temperature Gas-Cooled Reactor (HTGR) are discussed, and the development of interim mechanical properties and supporting data...

  16. Application of Hastelloy X in gas-cooled reactor systems

    Energy Technology Data Exchange (ETDEWEB)

    Brinkman, C.R.; Rittenhouse, P.L.; Corwin, W.R.; Strizak, J.P.; Lystrup, A.; DiStefano, J.R.

    1976-10-01

    Hastelloy X, an Ni--Cr--Fe--Mo alloy, may be an important structural alloy for components of gas-cooled reactor systems. Expected applications of this alloy in the High-Temperature Gas-Cooled Reactor (HTGR) are discussed, and the development of interim mechanical properties and supporting data are reported. Properties of concern include tensile, creep, creep-rupture, fatigue, creep-fatigue interaction, subcritical crack growth, thermal stability, and the influence of helium environments with controlled amounts of impurities on these properties. In order to develop these properties in helium environments that are expected to be prototypic of HTGR operating conditions, it was necessary to construct special environmental test systems. Details of construction and operating parameters are described. Interim results from tests designed to determine the above properties are presented. To date a fairly extensive amount of information has been generated on this material at Oak Ridge National Laboratory and elsewhere concerning behavior in air, which is reviewed. However, only limited data are available from tests conducted in helium. Comparisons of the fatigue and subcritical growth behavior in air between Hastelloy X and a number of other structural alloys are given.

  17. Performance of Air-cooled Engine Cylinders Using Blower Cooling

    Science.gov (United States)

    Schey, Oscar W; Ellerbrock, Herman H , Jr

    1936-01-01

    An investigation was made to obtain information on the minimum quantity of air and power required to cool conventional air cooled cylinders at various operating conditions when using a blower. The results of these tests show that the minimum power required for satisfactory cooling with an overall blower efficiency of 100 percent varied from 2 to 6 percent of the engine power depending on the operating conditions. The shape of the jacket had a large effect on the cylinder temperatures. Increasing the air speed over the front of the cylinder by keeping the greater part of the circumference of the cylinder covered by the jacket reduced the temperatures over the entire cylinder.

  18. Thermotechnical performance of an air-cooled tuyere with air cooling channels in series

    Science.gov (United States)

    Shen, Yuansheng; Zhou, Yuanyuan; Zhu, Tao; Duan, Guangbin

    2017-01-01

    To reduce the cooling air consumption for an air-cooled tuyere, an air-cooled tuyere with air cooling channels in series is developed based on several hypotheses, i.e., a transparent medium in the blast furnace, among others, and the related mathematical models are introduced and developed. Referring to the data from a BF site, the thermotechnical computation for the air-cooled tuyere was performed, and the results show that when the temperature of the inlet cooling air increases, the temperatures for the outlet cooling air, the outer surface of the tuyere, the walls of the air cooling channels and the center channel as well as the heat going into the center channel increase, but the heat absorbed by the cooling air flowing through the air cooling channels decreases. When the cooling air flow rate under the standard state increases, the physical parameters mentioned above change in an opposite directions. Compared to a water-cooled tuyere, the energy savings for an air-cooled tuyere are more than 0.23 kg/min standard coal.

  19. High Temperature Gas-Cooled Test Reactor Options Status Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-08-01

    Preliminary scoping calculations are being performed for a 100 MWt gas-cooled test reactor. The initial design uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to identify some reactor design features to investigate further. Current status of the effort is described.

  20. Coupled reactor kinetics and heat transfer model for heat pipe cooled reactors

    Science.gov (United States)

    Wright, Steven A.; Houts, Michael

    2001-02-01

    Heat pipes are often proposed as cooling system components for small fission reactors. SAFE-300 and STAR-C are two reactor concepts that use heat pipes as an integral part of the cooling system. Heat pipes have been used in reactors to cool components within radiation tests (Deverall, 1973); however, no reactor has been built or tested that uses heat pipes solely as the primary cooling system. Heat pipe cooled reactors will likely require the development of a test reactor to determine the main differences in operational behavior from forced cooled reactors. The purpose of this paper is to describe the results of a systems code capable of modeling the coupling between the reactor kinetics and heat pipe controlled heat transport. Heat transport in heat pipe reactors is complex and highly system dependent. Nevertheless, in general terms it relies on heat flowing from the fuel pins through the heat pipe, to the heat exchanger, and then ultimately into the power conversion system and heat sink. A system model is described that is capable of modeling coupled reactor kinetics phenomena, heat transfer dynamics within the fuel pins, and the transient behavior of heat pipes (including the melting of the working fluid). This paper focuses primarily on the coupling effects caused by reactor feedback and compares the observations with forced cooled reactors. A number of reactor startup transients have been modeled, and issues such as power peaking, and power-to-flow mismatches, and loading transients were examined, including the possibility of heat flow from the heat exchanger back into the reactor. This system model is envisioned as a tool to be used for screening various heat pipe cooled reactor concepts, for designing and developing test facility requirements, for use in safety evaluations, and for developing test criteria for in-pile and out-of-pile test facilities. .

  1. Nanostructured Catalytic Reactors for Air Purification Project

    Data.gov (United States)

    National Aeronautics and Space Administration — This SBIR Phase I project proposes the development of lightweight compact nanostructured catalytic reactors for air purification from toxic gaseous organic...

  2. Nanostructured Catalytic Reactors for Air Purification Project

    Data.gov (United States)

    National Aeronautics and Space Administration — This SBIR Phase II project proposes the development of lightweight compact nanostructured catalytic reactors for air purification from toxic gaseous organic...

  3. An Innovative Reactor Technology to Improve Indoor Air Quality

    Energy Technology Data Exchange (ETDEWEB)

    Rempel, Jane [TIAX LLC., Lexington, MA (United States)

    2013-03-30

    As residential buildings achieve tighter envelopes in order to minimize energy used for space heating and cooling, accumulation of indoor air pollutants such as volatile organic compounds (VOCs), becomes a major concern causing poor air quality and increased health risks. Current VOC removal methods include sorbents, ultraviolet photocatalytic oxidation (UVPCO), and increased ventilation, but these methods do not capture or destroy all VOCs or are prohibitively expensive to implement. TIAX's objective in this program was to develop a new VOC removal technology for residential buildings. This novel air purification technology is based on an innovative reactor and light source design along with UVPCO properties of the chosen catalyst to purify indoor air and enhance indoor air quality (IAQ). During the program we designed, fabricated and tested a prototype air purifier to demonstrate its feasibility and effectiveness. We also measured kinetics of VOC destruction on photocatalysts, providing deep insight into reactor design.

  4. Cooling Performance of Natural Circulation for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Suki; Chun, J. H.; Yum, S. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    This paper deals with the core cooling performance by natural circulation during normal operation and a flow channel blockage event in an open tank-in-pool type research reactor. The cooling performance is predicted by using the RELAP5/ MOD3.3 code. The core decay heat is usually removed by natural circulation to the reactor pool water in open tank-in-pool type research reactors with the thermal power less than several megawatts. Therefore, these reactors have generally no active core cooling system against a loss of normal forced flow. In reactors with the thermal power less than around one megawatt, the reactor core can be cooled down by natural circulation even during normal full power operation. The cooling performance of natural circulation in an open tank-in-pool type research reactor has been investigated during the normal natural circulation and a flow channel blockage event. It is found that the maximum powers without void generation at the hot channel are around 1.16 MW and 820 kW, respectively, for the normal natural circulation and the flow channel blockage event.

  5. Stirling Air Conditioner for Compact Cooling

    Energy Technology Data Exchange (ETDEWEB)

    None

    2010-09-01

    BEETIT Project: Infinia is developing a compact air conditioner that uses an unconventional high efficient Stirling cycle system (vs. conventional vapor compression systems) to produce cool air that is energy efficient and does not rely on polluting refrigerants. The Stirling cycle system is a type of air conditioning system that uses a motor with a piston to remove heat to the outside atmosphere using a gas refrigerant. To date, Stirling systems have been expensive and have not had the right kind of heat exchanger to help cool air efficiently. Infinia is using chip cooling technology from the computer industry to make improvements to the heat exchanger and improve system performance. Infinia’s air conditioner uses helium gas as refrigerant, an environmentally benign gas that does not react with other chemicals and does not burn. Infinia’s improvements to the Stirling cycle system will enable the cost-effective mass production of high-efficiency air conditioners that use no polluting refrigerants.

  6. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    Baldev Raj

    2009-06-01

    The paper gives an insight into basic as well as applied research being carried out at the Indira Gandhi Centre for Atomic Research for the development of advanced materials for sodium cooled fast reactors towards extending the life of reactors to nearly 100 years and the burnup of fuel to 2,00,000 MWd/t with an objective of providing fast reactor electricity at an affordable and competitive price.

  7. International working group on gas-cooled reactors. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    1981-01-15

    The purpose of the meeting was to provide a forum for exchange of information on safety and licensing aspects for gas-cooled reactors in order to provide comprehensive review of the present status and of directions for future applications and development. Contributions were made concerning the operating experience of the Fort St. Vrain (FSV) HTGR Power Plant in the United States of America, the experimental power station Arbeitsgemeinschaft Versuchsreaktor (AVR) in the Federal Republic of Germany, and the CO/sub 2/-cooled reactors in the United Kingdom such as Hunterson B and Hinkley Point B. The experience gained at each of these reactors has proved the high safety potential of Gas-cooled Reactor Power Plants.

  8. Computational Fluid Dynamics Analysis of Very High Temperature Gas-Cooled Reactor Cavity Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Frisani, Angelo; Hassan, Yassin A; Ugaz, Victor M

    2010-11-02

    The design of passive heat removal systems is one of the main concerns for the modular very high temperature gas-cooled reactors (VHTR) vessel cavity. The reactor cavity cooling system (RCCS) is a key heat removal system during normal and off-normal conditions. The design and validation of the RCCS is necessary to demonstrate that VHTRs can survive to the postulated accidents. The computational fluid dynamics (CFD) STAR-CCM+/V3.06.006 code was used for three-dimensional system modeling and analysis of the RCCS. A CFD model was developed to analyze heat exchange in the RCCS. The model incorporates a 180-deg section resembling the VHTR RCCS experimentally reproduced in a laboratory-scale test facility at Texas A&M University. All the key features of the experimental facility were taken into account during the numerical simulations. The objective of the present work was to benchmark CFD tools against experimental data addressing the behavior of the RCCS following accident conditions. Two cooling fluids (i.e., water and air) were considered to test the capability of maintaining the RCCS concrete walls' temperature below design limits. Different temperature profiles at the reactor pressure vessel (RPV) wall obtained from the experimental facility were used as boundary conditions in the numerical analyses to simulate VHTR transient evolution during accident scenarios. Mesh convergence was achieved with an intensive parametric study of the two different cooling configurations and selected boundary conditions. To test the effect of turbulence modeling on the RCCS heat exchange, predictions using several different turbulence models and near-wall treatments were evaluated and compared. The comparison among the different turbulence models analyzed showed satisfactory agreement for the temperature distribution inside the RCCS cavity medium and at the standpipes walls. For such a complicated geometry and flow conditions, the tested turbulence models demonstrated that the

  9. Fuel development for gas-cooled fast reactors

    Science.gov (United States)

    Meyer, M. K.; Fielding, R.; Gan, J.

    2007-09-01

    The Generation IV Gas-cooled Fast Reactor (GFR) concept is proposed to combine the advantages of high-temperature gas-cooled reactors (such as efficient direct conversion with a gas turbine and the potential for application of high-temperature process heat), with the sustainability advantages that are possible with a fast-spectrum reactor. The latter include the ability to fission all transuranics and the potential for breeding. The GFR is part of a consistent set of gas-cooled reactors that includes a medium-term Pebble Bed Modular Reactor (PBMR)-like concept, or concepts based on the Gas Turbine Modular Helium Reactor (GT-MHR), and specialized concepts such as the Very High-Temperature Reactor (VHTR), as well as actinide burning concepts [A Technology Roadmap for Generation IV Nuclear Energy Systems, US DOE Nuclear Energy Research Advisory Committee and the Generation IV International Forum, December 2002]. To achieve the necessary high power density and the ability to retain fission gas at high temperature, the primary fuel concept proposed for testing in the United States is dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the US and internationally include coated particle beds, ceramic clad fuel pins, and novel ceramic 'honeycomb' structures. Both mixed carbide and mixed nitride-based solid solutions are considered as fuel phases.

  10. High power density reactors based on direct cooled particle beds

    Science.gov (United States)

    Powell, J. R.; Horn, F. L.

    Reactors based on direct cooled High Temperature Gas Cooled Reactor (HTGR) type particle fuel are described. The small diameter particle fuel is packed between concentric porous cylinders to make annular fuel elements, with the inlet coolant gas flowing inwards. Hot exit gas flows out along the central channel of each element. Because of the very large heat transfer area in the packed beds, power densities in particle bed reactors (PBRs) are extremely high resulting in compact, lightweight systems. Coolant exit temperatures are high, because of the ceramic fuel temperature capabilities, and the reactors can be ramped to full power and temperature very rapidly. PBR systems can generate very high burst power levels using open cycle hydrogen coolant, or high continuous powers using closed cycle helium coolant. PBR technology is described and development requirements assessed.

  11. Development of Air-cooled Engines with Blower Cooling

    Science.gov (United States)

    Lohner, Kurt

    1933-01-01

    With the aid of a heating device, the heat transfer to cylinders with conical fins of various forms is determined both for shrouded and exposed cylinders. Simultaneously the pressure drop for overcoming the resistance to the motion of air between the fins of the enclosed cylinder is measured. Thus the relations between the heat transfer and the energy required for cooling are discovered. The investigations show that the heat transfer in a conducted air flow is much greater than in a free current and that further improvement, as compared with free exposure, is possible through narrower spaces between the fins.

  12. High-Temperature Gas-Cooled Test Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Laboratory; Bayless, Paul David [Idaho National Laboratory; Nelson, Lee Orville [Idaho National Laboratory; Gougar, Hans David [Idaho National Laboratory; Kinsey, James Carl [Idaho National Laboratory; Strydom, Gerhard [Idaho National Laboratory; Kumar, Akansha [Idaho National Laboratory

    2016-04-01

    A point design has been developed for a 200 MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched UCO fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technological readiness level, licensing approach and costs.

  13. Design of an Actinide Burning, Lead-Bismuth Cooled Reactor That Produces Low Cost Electricity

    Energy Technology Data Exchange (ETDEWEB)

    C. Davis; S. Herring; P. MacDonald; K. McCarthy; V. Shah; K. Weaver (INEEL); J. Buongiorno; R. Ballinger; K. Doyoung; M. Driscoll; P. Hejzler; M. Kazimi; N. Todreas (MIT)

    1999-07-01

    The purpose of this project is to investigate the suitability of lead-bismuth cooled fast reactors for producing low-cost electricity as well as for actinide burning. The goal is to identify and analyze the key technical issues in core neutronics, materials, thermal-hydraulics, fuels, and economics associated with the development of this reactor concept. The choice of lead-bismuth for the reactor coolant is an actinide burning fast reactor offers enhanced safety and reliability. The advantages of lead-bismuth over sodium as a coolant are related to the following material characteristics: chemical inertness with air and water; higher atomic number; lower vapor pressure at operating temperatures; and higher boiling temperature. Given the status of the field, it was agreed that the focus of this investigation in the first two years will be on the assessment of approaches to optimize core and plant arrangements in order to provide maximum safety and economic potential in this type of reactor.

  14. Capital cost: gas cooled fast reactor plant

    Energy Technology Data Exchange (ETDEWEB)

    1977-09-01

    The results of an investment cost study for a 900 MW(e) GCFR central station power plant are presented. The capital cost estimate arrived at is based on 1976 prices and a conceptual design only, not a mature reactor design.

  15. Simple analysis of an External Vessel Cooling Thermosyphon for a Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jae Young; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Song, Sub Lee [Handong Global University, Pohang (Korea, Republic of)

    2015-05-15

    KALIMER has three different DHR systems: two non-safety grade systems and one safety grade system. The non-safety grade systems are an IRACS (Intermediate Reactor Auxiliary Cooling System) and a steam/feedwater system. The safety grade system is a PDRC (Passive Decay Heat Removal Circuit). In case of the foreign reactor designs, ABTR (Advanced Burner Test Reactor) has a DRACS (Direct Reactor Auxiliary Cooling System), a PFBR (Indian Prototype Fast Breeder Reactor) has an SGDHRS (Safety Grade Decay Heat Removal System), and an EFR (European Fast Reactor) has DRC (Direct Reactor Cooling). Those designs have advantage on relatively high decay heat removal capacity. However, larger vessel size due to subsidiary in-vessel structure and possible accident propagation to reactor induced by sodium fire. In this paper, an ex-vessel thermosyphon design was proposed for the removal of decay heat for an iSFR. The proposed ex-vessel thermosyphon was designed to remove decay heat in both transient cases and BDBA cases, such as vessel failure. Proper working fluid was selected based on thermodynamic properties and chemical stability. Mercury was chosen as the working fluid, and SUS 314 was used for the corresponding structure material. Possible chemical reactions and adverse effects from using the thermosyphon were inherently eliminated by the system layout. A model for a high-temperature thermosyphon and numerical algorithms were used for the analysis. As a result of the simulation, the thermosyphon design was optimized, and it showed sufficient DHR performance to maintain core integrity.

  16. Gas-cooled reactors: the importance of their development

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.

    1979-06-01

    The nearest term GCR is the steam-cycle HTGR, which can be used for both power and process steam production. Use of SC-HTGRs permits timely introduction of thorium fuel cycles and of high-thermal-efficiency reactors, decreasing the need for mined U/sub 3/O/sub 8/ before arrival of symbiotic fueling of fast-thermal reactor systems. The gas-turbine HTGR offers prospects of lower capital costs than other nuclear reactors, but it appears to require longer and more costly development than the SC-HTGR. Accelerated development of the GT-HTGR is needed to gain the advantages of timely introduction. The Gas-Cooled Fast Breeder Reactor (GCFR) offers the possibility of fast breeder reactors with lower capital costs and with higher breeding ratios from oxide fuels. The VHTR provides high-temperature heat for hydrogen production.

  17. Simulation of Reactor Transient and Design Criteria of Sodium-cooled Fast Reactors

    OpenAIRE

    Gottfridsson, Filip

    2010-01-01

    The need for energy is growing in the world and the market of nuclear power is now once more expanding. Some issues of the current light-water reactors can be solved by the next generation of nuclear power, Generation IV, where sodium-cooled reactors are one of the candidates. Phénix was a French prototype sodium-cooled reactor, which is seen as a success. Although it did encounter an earlier unexperienced phenomenon, A.U.R.N., in which a negative reactivity transient followed by an oscillati...

  18. Control rod drive for high temperature gas cooled reactor

    Institute of Scientific and Technical Information of China (English)

    DengJun-Xian; XuJi-Ming; 等

    1998-01-01

    This control rod drive is developed for HTR-10 high temperature gas cooled test reactor.The stepmotor is prefered to improve positioning of the control rod and the scram behavior.The preliminary test in 1600170 ambient temperature shows that the selected stepmotor and transmission system can meet the main operation function requirements of HTR-10.

  19. Thermoelectric Air Cooling For Cars

    Directory of Open Access Journals (Sweden)

    Manoj S. Raut

    2012-05-01

    Full Text Available India is the second most populous country in the world with over 1.21billion people (estimated for April, 2011,more than sixth of the world’s population. India is projected to be world’s most populous country by 2025,surpassing china, its population exceeding 1.6 billion people by 2050.Comparing with the population there are 2.65 million cars sold in India as of march 2011.According to the society of Indian automotive manufacturer, annual car sales are projected to increase up to 5 million vehicles by 2015 and more than a 9 million by 2020.By 2050,the country is expected to top of the world in car volumes with approximately 611 million vehicles on the nation’s roads.The above data shows that, as the population increase the no. of vehicles also increase. Today, an automobile is a necessity for everyone. For a long or short journey people need car regard to thesafety, environment and most important comfort. Owing to these reasons, many vehicles are equipped with heating, ventilating and air conditioning system. In today’s world, no one feel comfortable in a vehicle without HVAC system. Therefore, HVAC becomes an integral part of human life. Today’s present HVAC system is very efficient and reliable but it has some demerits. It has been observed during the last two decades that the O3 –layer is slowly destroyed because of the refrigerant (CFC and HFC used for the refrigeration and air –conditioning purposes. The common refrigerant used is HFC’s which are leaked and slowly climb into the atmosphere. When they reach to O3 layer they act on O3 –molecules and the layer of O3 is destroyed.

  20. The Effect of Duct Level on the Performance of Reactor Vault Cooling System in the PGSFR

    Energy Technology Data Exchange (ETDEWEB)

    Yeom, Sujin; Ryu, Seung Ho; Kim, Dehee; Lee, Tae-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Development of the prototype gen-Ⅵ sodium-cooled fast reactor (PGSFR) has been ongoing in Korea Atomic Energy Research Institute (KAERI). A reactor vault cooling system (RVCS), one of passive decay heat removal systems (PDHRS), passively removes core decay heat by chimney effect when severe accidents occur. The air cooling path is located around containment vessel (CV). An air separator which divides the downstream air and the upstream air is installed between CV and the concrete wall. To design the RVCS, key design parameters such as stack height, gap size between the concrete wall and the air separator, gap size between the air separator and the CV, thickness and layer composition of the air separator have to be determined. A duct level is one of these design parameters. It denotes the height of the upstream air path and related to the heat transfer length from CV to air. The duct level should be optimized with considering structural reliability and heat removal performance. Thus, in this paper, the heat removal performance of RVCS is evaluated depends on the duct level using 1D system design code, that is developed by KAERI autonomously, and commercial CFD program for optimum design of RVCS In this paper, the heat removal performance of RVCS is evaluated depends on the duct level using PARS2- LMR code and commercial CFD program for optimum design of RVCS to satisfy both conflicting needs, structural reliability and cooling performance. As a result of PARS2-LMR code analysis, it was observed that the heat removal rate increases as increase of duct level and the geometrical conditions, that satisfy the design limitations, were obtained. To qualitatively observe the trends of local temperature distribution, CFD simulations were conducted and hotspots were observed at the upper region of ducts for the low duct level case.

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

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

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

  2. Enhancing VHTR Passive Safety and Economy with Thermal Radiation Based Direct Reactor Auxiliary Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Hongbin Zhang; Ling Zou; Xiaodong Sun

    2012-06-01

    One of the most important requirements for Gen. IV Very High Temperature Reactor (VHTR) is passive safety. Currently all the gas cooled version of VHTR designs use Reactor Vessel Auxiliary Cooling System (RVACS) for passive decay heat removal. The decay heat first is transferred to the core barrel by conduction and radiation, and then to the reactor vessel by thermal radiation and convection; finally the decay heat is transferred to natural circulated air or water systems. RVACS can be characterized as a surface based decay heat removal system. The RVACS is especially suitable for smaller power reactors since small systems have relatively larger surface area to volume ratio. However, RVACS limits the maximum achievable power level for modular VHTRs due to the mismatch between the reactor power (proportional to volume) and decay heat removal capability (proportional to surface area). When the relative decay heat removal capability decreases, the peak fuel temperature increases, even close to the design limit. Annular core designs with inner graphite reflector can mitigate this effect; therefore can further increase the reactor power. Another way to increase the reactor power is to increase power density. However, the reactor power is also limited by the decay heat removal capability. Besides the safety considerations, VHTRs also need to be economical in order to compete with other reactor concepts and other types of energy sources. The limit of decay heat removal capability set by using RVACS has affected the economy of VHTRs. A potential alternative solution is to use a volume-based passive decay heat removal system, called Direct Reactor Auxiliary Cooling Systems (DRACS), to remove or mitigate the limitation on decay heat removal capability. DRACS composes of natural circulation loops with two sets of heat exchangers, one on the reactor side and another on the environment side. For the reactor side, cooling pipes will be inserted into holes made in the outer or

  3. Preliminary Assessment of a Debris Bed Cooling Performance for Demonstration Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Chung Ho; Park, Chang Gyu; Song, Hoon; Kim, Young Gyun; Jeong, Hae Yong; Chang, Jin Wook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-05-15

    In the case of the sodium-cooled fast reactor such as KALIMER-600, Hypothetical Core Disruptive Accident (HCDA) attributed from mass nuclear fuel melting is unlikely to occur due to defense in depth concepts to meet requirements of redundancy and diversity. Multiple faults such as loss of flow, loss of heat sink, or transient overpower without scram are to lead rising the power level until cladding failure as reactivity increasing. The fact that metallic fuel melts at a lower temperature than the cladding allows significant in-pin- fuel motion to occur prior to cladding failure. Also, the combination of Doppler and axial expansion feedback and negative feedback associated with the in-pin fuel relocation prevents the reactivity from reaching prompt critical. Finally, the resulting reactivity and power reductions help prevent fuel temperatures from rising more than the fuel melting temperature. It is more difficult to occur HCDA in a metallic fueled core because reactor power and heat removal capability is maintained in balance by inherent safety characteristics However, for the future design of sodium-cooled fast reactor, the evaluation of the safety performance and the determination of containment requirements may be worth considering due to the triple-fault accident sequences of extremely low probability of occurrence that leads to core melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will be required as a design requirement for the future design of sodium cooled fast reactor. Also, proof of the capacity of the debris bed cooling is an essential condition to solve the problem of in-vessel retention of the core debris. Accordingly, evaluation of a packed debris bed cooling performance with single phase flow for demonstration sodium-cooled fast reactor was carried out for proof of the in-vessel retention of the core debris

  4. System Study: Reactor Core Isolation Cooling 1998-2014

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-12-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  5. System Study: Reactor Core Isolation Cooling 1998–2013

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-01-31

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2013 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  6. System Study: Reactor Core Isolation Cooling 1998–2012

    Energy Technology Data Exchange (ETDEWEB)

    T. E. Wierman

    2013-10-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2012 for selected components were obtained from the Equipment Performance and Information Exchange (EPIX). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing trend was identified in the HPCI results. Statistically significant decreasing trends were identified for RCIC start-only and 8-hour trends.

  7. Anomalous Effects in Air While Cooling Water

    CERN Document Server

    Sardo, Rachel

    2008-01-01

    Water is a unique compound with many anomalies and properties not fully understood. Designing an experiment in the laboratory to study such anomalies, we set up a series of experiments where a tube was placed inside a sealed container with thermocouples attached to the outer surface of the tube and in the air adjacent to the tube. Alternately, deionized water and other compounds were added to the tube and cooled to freezing. Several of the thermocouples suspended in the air and adjacent to the tube showed thermal oscillations as the overall temperature of the container was decreasing. The temperature of the thermocouples increased and decreased in a sinusoidal way during part of the cool down to freezing. Thermal oscillations as large as 3 degrees Celsius were recorded with typical frequencies of about 5 oscillations per minute.

  8. On the Method of Air Jet Cooling in Green Manufacturing

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    Green cooling is an important technology in green manufacturing. In the way of jetting, cooling airflow is used in the experiments of metal material cutting, by compari- son of the changes of some technological factors, such as cutting heat, surface finish, in the process of jet cooling, pour cooling and natural cooling, we can draw the conclusion that air jet cooling has a better cooling effect and green function. It can be widely used in both aditional and automatic green manufacturing.

  9. Superheated Water-Cooled Small Modular Underwater Reactor Concept

    Directory of Open Access Journals (Sweden)

    Koroush Shirvan

    2016-12-01

    Full Text Available A novel fully passive small modular superheated water reactor (SWR for underwater deployment is designed to produce 160 MWe with steam at 500ºC to increase the thermodynamic efficiency compared with standard light water reactors. The SWR design is based on a conceptual 400-MWe integral SWR using the internally and externally cooled annular fuel (IXAF. The coolant boils in the external channels throughout the core to approximately the same quality as a conventional boiling water reactor and then the steam, instead of exiting the reactor pressure vessel, turns around and flows downward in the central channel of some IXAF fuel rods within each assembly and then flows upward through the rest of the IXAF pins in the assembly and exits the reactor pressure vessel as superheated steam. In this study, new cladding material to withstand high temperature steam in addition to the fuel mechanical and safety behavior is investigated. The steam temperature was found to depend on the thermal and mechanical characteristics of the fuel. The SWR showed a very different transient behavior compared with a boiling water reactor. The inter-play between the inner and outer channels of the IXAF was mainly beneficial except in the case of sudden reactivity insertion transients where additional control consideration is required.

  10. Floor cooling and air-cooling, the effects on thermal comfort or different cooling systems

    Energy Technology Data Exchange (ETDEWEB)

    Sijpheer, N.C.; Bakker, E.J.; Ligthart, F.A.T.M.; Opstelten, I.J. [ECN Energie in de Gebouwde Omgeving en Netten, Petten (Netherlands)

    2007-09-15

    One of the research areas of the Energy research Centre of the Netherlands (ECN) concerns the built environment. Several facilities to conduct research activities are at ECN's disposal. One of these facilities, are five research dwellings located on the premises of ECN. Measured data from these facilities together with weather data and computer models are used to evaluate innovative energy concepts and components in energy systems. Experiments with different cooling systems in ECN's research dwellings are executed to evaluate their effective influence on both energy use and thermal comfort. Influence of inhabitants' behaviour is taken into account in these experiments. The thermal comfort is indicated by the Predicted Mean Vote (PMV) as defined by P.O. Fanger. For this paper, the results of measurements with a floor cooling and air cooling system are assessed. Effects on the PMV measured during experiments with the two different cooling systems will be presented.

  11. Convective cooling in a pool-type research reactor

    Science.gov (United States)

    Sipaun, Susan; Usman, Shoaib

    2016-01-01

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

  12. Convective cooling in a pool-type research reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-22

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

  13. Comparison of Air Cooled and Evaporatively Cooled Refrigerartion Systems – A Review Paper

    Directory of Open Access Journals (Sweden)

    V. V. Birangane

    2014-06-01

    Full Text Available The air cooled condensers are widely used as they are less costly and give satisfactory performance. But their performance is greatly affected by the temperature of cooling media which is ambient air. To deal this problem we can use water cooled condenser. But their cost and maintenance limit their use. The performance improvement of Air cooled condensers can be achieved by using evaporative cooling. This method may prove quiet effective and less costly. There are researchers working on the above issue. Few of them have successfully implemented the research in practice. The paper deals with a few papers using the evaporative cooling. The applications include domestic as well as industrial.

  14. Accident analysis of heavy water cooled thorium breeder reactor

    Science.gov (United States)

    Yulianti, Yanti; Su'ud, Zaki; Takaki, Naoyuki

    2015-04-01

    Thorium has lately attracted considerable attention because it is accumulating as a by-product of large scale rare earth mining. The objective of research is to analyze transient behavior of a heavy water cooled thorium breeder that is designed by Tokai University and Tokyo Institute of Technology. That is oxide fueled, PWR type reactor with heavy water as primary coolant. An example of the optimized core has relatively small moderator to fuel volume ratio (MFR) of 0.6 and the characteristics of the core are burn-up of 67 GWd/t, breeding ratio of 1.08, burn-up reactivity loss during cycles of nuclear reactor accidents types examined here is Unprotected Transient over Power (UTOP) due to withdrawing of the control rod that result in the positive reactivity insertion so that the reactor power will increase rapidly. Another accident type is Unprotected Loss of Flow (ULOF) that caused by failure of coolant pumps. To analyze the reactor accidents, neutron distribution calculation in the nuclear reactor is the most important factor. The best expression for the neutron distribution is the Boltzmann transport equation. However, solving this equation is very difficult so that the space-time diffusion equation is commonly used. Usually, space-time diffusion equation is solved by employing a point kinetics approach. However, this approach is less accurate for a spatially heterogeneous nuclear reactor and the nuclear reactor with quite large reactivity input. Direct method is therefore used to solve space-time diffusion equation which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference method is solved by using iterative methods. The indication of UTOP accident is decreasing macroscopic absorption cross-section that results large external reactivity, and ULOF accident is indicated by decreasing coolant flow. The power reactor has a peak value before reactor has new balance condition

  15. Specific power of liquid-metal-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dobranich, D.

    1987-10-01

    Calculations of the core specific power for conceptual space-based liquid-metal-cooled reactors, based on heat transfer considerations, are presented for three different fuel types: (1) pin-type fuel; (2) cermet fuel; and (3) thermionic fuel. The calculations are based on simple models and are intended to provide preliminary comparative results. The specific power is of interest because it is a measure of the core mass required to produce a given amount of power. Potential problems concerning zero-g critical heat flux and loss-of-coolant accidents are also discussed because these concerns may limit the core specific power. Insufficient experimental data exists to accurately determine the critical heat flux of liquid-metal-cooled reactors in space; however, preliminary calculations indicate that it may be a concern. Results also indicate that the specific power of the pin-type fuels can be increased significantly if the gap between the fuel and the clad is eliminated. Cermet reactors offer the highest specific power because of the excellent thermal conductivity of the core matrix material. However, it may not be possible to take fuel advantage of this characteristic when loss-of-coolant accidents are considered in the final core design. The specific power of the thermionic fuels is dependent mainly on the emitter temperature. The small diameter thermionic fuels have specific powers comparable to those of pin-type fuels. 11 refs., 12 figs, 2 tabs.

  16. Delayed gamma power measurement for sodium-cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Coulon, R., E-mail: romain.coulon@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Normand, S., E-mail: stephane.normand@cea.f [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Ban, G., E-mail: ban@lpccaen.in2p3.f [ENSICAEN, 6 Boulevard Marechal Juin, F-14050 Caen Cedex 4 (France); Barat, E.; Montagu, T.; Dautremer, T. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Brau, H.-P. [ICSM, Centre de Marcoule, BP 17171 F-30207 Bagnols sur Ceze (France); Dumarcher, V. [AREVA NP, SET, F-84500 Bollene (France); Michel, M.; Barbot, L.; Domenech, T.; Boudergui, K.; Bourbotte, J.-M. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France); Jousset, P. [CEA, LIST, Departement des Capteurs, du Signal et de l' Information, F-91191 Gif-sur-Yvette (France); Barouch, G.; Ravaux, S.; Carrel, F. [CEA, LIST, Laboratoire Modelisation Simulation et Systemes, F-91191 Gif-sur-Yvette (France); Saurel, N. [CEA, DAM, Laboratoire Mesure de Dechets et Expertise, F-21120 Is-sur-Tille (France); Frelin-Labalme, A.-M.; Hamrita, H. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, F-91191 Gif-sur-Yvette (France)

    2011-01-15

    Graphical abstract: Display Omitted Research highlights: {sup 20}F and {sup 23}Ne tagging agents are produced by fast neutron flux. {sup 20}F signal has been measured at the SFR Phenix prototype. A random error of only 3% for an integration time of 2 s could be achieved. {sup 20}F and {sup 23}Ne power measurement has a reduced temperature influence. Burn-up impact could be limited by simultaneous {sup 20}F and {sup 23}Ne measurement. - Abstract: Previous works on pressurized water reactors show that the nitrogen 16 activation product can be used to measure thermal power. Power monitoring using a more stable indicator than ex-core neutron measurements is required for operational sodium-cooled fast reactors, in order to improve their economic efficiency at the nominal operating point. The fluorine 20 and neon 23 produced by (n,{alpha}) and (n,p) capture in the sodium coolant have this type of convenient characteristic, suitable for power measurements with low build-up effects and a potentially limited temperature, flow rate, burn-up and breeding dependence. This method was tested for the first time during the final tests program of the French Phenix sodium-cooled fast reactor at CEA Marcoule, using the ADONIS gamma pulse analyzer. Despite a non-optimal experimental configuration for this application, the delayed gamma power measurement was pre-validated, and found to provide promising results.

  17. Improving fuel cycle design and safety characteristics of a gas cooled fast reactor

    NARCIS (Netherlands)

    van Rooijen, W.F.G.

    2006-01-01

    This research concerns the fuel cycle and safety aspects of a Gas Cooled Fast Reactor, one of the so-called "Generation IV" nuclear reactor designs. The Generation IV Gas Cooled Fast Reactor uses helium as coolant at high temperature. The goal of the GCFR is to obtain a "closed nuclear fuel cycle",

  18. COMPARISON OF COOLING SCHEMES FOR HIGH HEAT FLUX COMPONENTS COOLING IN FUSION REACTORS

    Directory of Open Access Journals (Sweden)

    Phani Kumar Domalapally

    2015-04-01

    Full Text Available Some components of the fusion reactor receives high heat fluxes either during the startup and shutdown or during the operation of the machine. This paper analyzes different ways of enhancing heat transfer using helium and water for cooling of these high heat flux components and then conclusions are drawn to decide the best choice of coolant, for usage in near and long term applications.

  19. Shape optimization of a sodium cooled fast reactor

    Science.gov (United States)

    Schmitt, Damien; Allaire, Grégoire; Pantz, Olivier; Pozin, Nicolas

    2014-06-01

    Traditional designs of sodium cooled fast reactors have a positive sodium expansion feedback. During a loss of flow transient without scram, sodium heating and boiling thus insert a positive reactivity and prevents the power from decreasing. Recent studies led at CEA, AREVA and EDF show that cores with complex geometries can feature a very low or even a negative sodium void worth.(1, 2) Usual optimization methods for core conception are based on a parametric description of a given core design(3).(4) New core concepts and shapes can then only be found by hand. Shape optimization methods have proven very efficient in the conception of optimal structures under thermal or mechanical constraints.(5, 6) First studies show that these methods could be applied to sodium cooled core conception.(7) In this paper, a shape optimization method is applied to the conception of a sodium cooled fast reactor core with low sodium void worth. An objective function to be minimized is defined. It includes the reactivity change induced by a 1% sodium density decrease. The optimization variable is a displacement field changing the core geometry from one shape to another. Additionally, a parametric optimization of the plutonium content distribution of the core is made, so as to ensure that the core is kept critical, and that the power shape is flat enough. The final shape obtained must then be adjusted to a get realistic core layout. Its caracteristics can be checked with reference neutronic codes such as ERANOS. Thanks to this method, new shapes of reactor cores could be inferred, and lead to new design ideas.

  20. Startup of air-cooled condensers and dry cooling towers at low temperatures of the cooling air

    Science.gov (United States)

    Milman, O. O.; Ptakhin, A. V.; Kondratev, A. V.; Shifrin, B. A.; Yankov, G. G.

    2016-05-01

    The problems of startup and performance of air-cooled condensers (ACC) and dry cooling towers (DCT) at low cooling air temperatures are considered. Effects of the startup of the ACC at sub-zero temperatures are described. Different options of the ACC heating up are analyzed, and examples of existing technologies are presented (electric heating, heating up with hot air or steam, and internal and external heating). The use of additional heat exchanging sections, steam tracers, in the DCT design is described. The need for high power in cases of electric heating and heating up with hot air is noted. An experimental stand for research and testing of the ACC startup at low temperatures is described. The design of the three-pass ACC unit is given, and its advantages over classical single-pass design at low temperatures are listed. The formation of ice plugs inside the heat exchanging tubes during the start-up of ACC and DCT at low cooling air temperatures is analyzed. Experimental data on the effect of the steam flow rate, steam nozzle distance from the heat-exchange surface, and their orientation in space on the metal temperature were collected, and test results are analyzed. It is noted that the surface temperature at the end of the heat up is almost independent from its initial temperature. Recommendations for the safe start-up of ACCs and DCTs are given. The heating flow necessary to sufficiently heat up heat-exchange surfaces of ACCs and DCTs for the safe startup is estimated. The technology and the process of the heat up of the ACC with the heating steam external supply are described by the example of the startup of the full-scale section of the ACC at sub-zero temperatures of the cooling air, and the advantages of the proposed start-up technology are confirmed.

  1. THATCH: A computer code for modelling thermal networks of high- temperature gas-cooled nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kroeger, P.G.; Kennett, R.J.; Colman, J.; Ginsberg, T. (Brookhaven National Lab., Upton, NY (United States))

    1991-10-01

    This report documents the THATCH code, which can be used to model general thermal and flow networks of solids and coolant channels in two-dimensional r-z geometries. The main application of THATCH is to model reactor thermo-hydraulic transients in High-Temperature Gas-Cooled Reactors (HTGRs). The available modules simulate pressurized or depressurized core heatup transients, heat transfer to general exterior sinks or to specific passive Reactor Cavity Cooling Systems, which can be air or water-cooled. Graphite oxidation during air or water ingress can be modelled, including the effects of added combustion products to the gas flow and the additional chemical energy release. A point kinetics model is available for analyzing reactivity excursions; for instance due to water ingress, and also for hypothetical no-scram scenarios. For most HTGR transients, which generally range over hours, a user-selected nodalization of the core in r-z geometry is used. However, a separate model of heat transfer in the symmetry element of each fuel element is also available for very rapid transients. This model can be applied coupled to the traditional coarser r-z nodalization. This report described the mathematical models used in the code and the method of solution. It describes the code and its various sub-elements. Details of the input data and file usage, with file formats, is given for the code, as well as for several preprocessing and postprocessing options. The THATCH model of the currently applicable 350 MW{sub th} reactor is described. Input data for four sample cases are given with output available in fiche form. Installation requirements and code limitations, as well as the most common error indications are listed. 31 refs., 23 figs., 32 tabs.

  2. Study on disposal method of graphite blocks and storage of spent fuel for modular gas-cooled reactor. Joint research

    Energy Technology Data Exchange (ETDEWEB)

    Sumita, Junya; Sawa, Kazuhiro; Kunitomi, Kazuhiko [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Tsuchie, Yasuo; Urakami, Masao [Japan Atomic Power Co., Tokyo (Japan)

    2003-02-01

    This report describes the result of study on disposal method of graphite blocks in future block-type reactor. Present study was carried out within a framework of joint research, ''Research of Modular High Temperature Gas-cooled Reactors (No. 3)'', between Japan Atomic Energy Research Institute (JAERI) and the Japan Atomic Power Company (JAPCO), in 2000. In this study, activities in fuel and reflector graphite blocks were evaluated and were compared with the disposal limits defined as low-level of radioactive waste. As a result, it was found that the activity for only C-14 was higher than disposal limits for the low-level of radioactive waste and that the amount of air in the graphite is important to evaluate precisely of C-14 activity. In addition, spent fuels can be stored in air-cooled condition at least after two years cooling in the storage pool. (author)

  3. CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

    OpenAIRE

    Pengcheng Zhao; Kangli Shi; Shuzhou Li; Jingchao Feng; Hongli Chen

    2016-01-01

    Small modular reactor (SMR) has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR) is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100) is being developed by University of Science and Technology of China (USTC). In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kineti...

  4. Experimental Studies of NGNP Reactor Cavity Cooling System With Water

    Energy Technology Data Exchange (ETDEWEB)

    Corradini, Michael; Anderson, Mark; Hassan, Yassin; Tokuhiro, Akira

    2013-01-16

    This project will investigate the flow behavior that can occur in the reactor cavity cooling system (RCCS) with water coolant under the passive cooling-mode of operation. The team will conduct separate-effects tests and develop associated scaling analyses, and provide system-level phenomenological and computational models that describe key flow phenomena during RCCS operation, from forced to natural circulation, single-phase flow and two-phase flow and flashing. The project consists of the following tasks: Task 1. Conduct separate-effects, single-phase flow experiments and develop scaling analyses for comparison to system-level computational modeling for the RCCS standpipe design. A transition from forced to natural convection cooling occurs in the standpipe under accident conditions. These tests will measure global flow behavior and local flow velocities, as well as develop instrumentation for use in larger scale tests, thereby providing proper flow distribution among standpipes for decay heat removal. Task 2. Conduct separate-effects experiments for the RCCS standpipe design as two-phase flashing occurs and flow develops. As natural circulation cooling continues without an ultimate heat sink, water within the system will heat to temperatures approaching saturation , at which point two-phase flashing and flow will begin. The focus is to develop a phenomenological model from these tests that will describe the flashing and flow stability phenomena. In addition, one could determine the efficiency of phase separation in the RCCS storage tank as the two-phase flashing phenomena ensues and the storage tank vents the steam produced. Task 3. Develop a system-level computational model that will describe the overall RCCS behavior as it transitions from forced flow to natural circulation and eventual two-phase flow in the passive cooling-mode of operation. This modeling can then be used to test the phenomenological models developed as a function of scale.

  5. Gas turbine vane cooling air insert

    Energy Technology Data Exchange (ETDEWEB)

    North, W.E.; Hultgren, K.G.; Dishman, C.D.; Van Heusden, G.S.

    1992-09-08

    This patent describes a gas turbine. It comprises turbine vanes, each of the vanes supplied with cooling air and having: an airfoil portion forming a first cavity having an insert disposed therein for directing the flow of the cooling air, the insert having first and second insert ends; a shroud portion from which the airfoil portion extends, the insert attached to the shroud portion at the first insert end; an insert extension extending through a portion of the insert and extending beyond the first insert end, the insert extension and the insert forming an annular gap therebetween separating the insert from the insert extension; a plate covering at least a portion of the shroud, the plate having a first hole formed therein through which the insert extension extends; and at least a first seal extending between the insert extension and the insert, and sealing the annular gap therebetween. This patent also describes a method of making a gas turbine. It comprises welding a first tubular insert adjacent its first end to a vane outer shroud; partially inserting a second tubular insert into the first tubular member and attaching the second tubular insert thereto; placing a plate having a hole formed therein on the outer shroud so that the hole surrounds the second tubular insert; and attaching the second tubular insert to the plate by placing a first seal between the first and second tubular inserts and attaching the first seal to each of the first and second tubular inserts, and placing a second seal between the second tubular insert and the plate and welding the second seal to the second tubular insert and the plate.

  6. Transient Load Following and Control Analysis of Advanced S-CO2 Power Conversion with Dry Air Cooling

    Energy Technology Data Exchange (ETDEWEB)

    Moisseytsev, Anton; Sienicki, James J.

    2016-01-01

    Supercritical carbon dioxide (S-CO2) Brayton cycles are under development as advanced energy converters for advanced nuclear reactors, especially the Sodium-Cooled Fast Reactor (SFR). The use of dry air cooling for direct heat rejection to the atmosphere ultimate heat sink is increasingly becoming a requirement in many regions due to restrictions on water use. The transient load following and control behavior of an SFR with an S-CO2 cycle power converter utilizing dry air cooling have been investigated. With extension and adjustment of the previously existing control strategy for direct water cooling, S-CO2 cycle power converters can also be used for load following operation in regions where dry air cooling is a requirement

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

    Energy Technology Data Exchange (ETDEWEB)

    Jensen, S.E.; Nonboel, E

    1999-05-01

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

  8. CFD Model Development and validation for High Temperature Gas Cooled Reactor Cavity Cooling System (RCCS) Applications

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Yassin [Univ. of Wisconsin, Madison, WI (United Texas A & M Univ., College Station, TX (United States); Corradini, Michael; Tokuhiro, Akira; Wei, Thomas Y.C.

    2014-07-14

    The Reactor Cavity Cooling Systems (RCCS) is a passive safety system that will be incorporated in the VTHR design. The system was designed to remove the heat from the reactor cavity and maintain the temperature of structures and concrete walls under desired limits during normal operation (steady-state) and accident scenarios. A small scale (1:23) water-cooled experimental facility was scaled, designed, and constructed in order to study the complex thermohydraulic phenomena taking place in the RCCS during steady-state and transient conditions. The facility represents a portion of the reactor vessel with nine stainless steel coolant risers and utilizes water as coolant. The facility was equipped with instrumentation to measure temperatures and flow rates and a general verification was completed during the shakedown. A model of the experimental facility was prepared using RELAP5-3D and simulations were performed to validate the scaling procedure. The experimental data produced during the steady-state run were compared with the simulation results obtained using RELAP5-3D. The overall behavior of the facility met the expectations. The facility capabilities were confirmed to be very promising in performing additional experimental tests, including flow visualization, and produce data for code validation.

  9. Steam-Reheat Option for Supercritical-Water-Cooled Reactors

    Science.gov (United States)

    Saltanov, Eugene

    SuperCritical-Water-cooled Reactors (SCWRs) are being developed as one of the Generation-IV nuclear-reactor concepts. Main objectives of the development are to increase thermal efficiency of a Nuclear Power Plant (NPP) and to decrease capital and operational costs. The first objective can be achieved by introducing nuclear steam reheat inside a reactor and utilizing regenerative feedwater heaters. The second objective can be achieved by designing a steam cycle that closely matches that of the mature supercritical fossil-fuelled power plants. The feasibility of these objectives is discussed. As a part of this discussion, heat-transfer calculations have been performed and analyzed for SuperCritical-Water (SCW) and SuperHeated-Steam (SHS) channels of the proposed reactor concept. In the calculations a uniform and three non-uniform Axial Heat Flux Profiles (AHFPs) were considered for six different fuels (UO2, ThO 2, MOX, UC2, UC, and UN) and at average and maximum channel power. Bulk-fluid, sheath, and fuel centerline temperatures as well as the Heat Transfer Coefficient (HTC) profiles were obtained along the fuel-channel length. The HTC values are within a range of 4.7--20 kW/m2·K and 9.7--10 kW/m2·K for the SCW and SHS channels respectively. The main conclusion is that while all the mentioned fuels may be used for the SHS channel, only UC2, UC, or UN are suitable for a SCW channel, because their fuel centerline temperatures are at least 1000°C below melting point, while that of UO2, ThO2 , and MOX may reach melting point.

  10. An Experimental Test Facility to Support Development of the Fluoride Salt Cooled High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoder Jr, Graydon L [ORNL; Aaron, Adam M [ORNL; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK); Fugate, David L [ORNL; Holcomb, David Eugene [ORNL; Kisner, Roger A [ORNL; Peretz, Fred J [ORNL; Robb, Kevin R [ORNL; Wilgen, John B [ORNL; Wilson, Dane F [ORNL

    2014-01-01

    The need for high-temperature (greater than 600 C) energy exchange and delivery systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The Fluoride Salt-Cooled High-Temperature Reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during the development process. The experimental facility described here has been constructed to support the development of the Fluoride Salt Cooled High Temperature Reactor concept. The facility is capable of operating at up to 700 C and incorporates a centrifugal pump to circulate FLiNaK salt through a removable test section. A unique inductive heating technique is used to apply heat to the test section, allowing heat transfer testing to be performed. An air-cooled heat exchanger removes added heat. Supporting loop infrastructure includes a pressure control system; trace heating system; and a complement of instrumentation to measure salt flow, temperatures, and pressures around the loop. The initial experiment is aimed at measuring fluoride salt heat transfer inside a heated pebble bed similar to that used for the core of the pebble bed advanced high-temperature reactor. This document describes the details of the loop design, auxiliary systems used to support the facility, the inductive heating system, and facility capabilities.

  11. Synthesis of ZnO particles in a quench-cooled flame reactor

    DEFF Research Database (Denmark)

    Hansen, Jens Peter; Jensen, Joakim Reimer; Livbjerg, Hans

    2001-01-01

    . At the highest tested production rate, the specific surface area of the ZnO particles increases from 20 to 60 m(2)/g when quenching is employed. The particles are characterized by BET surface area measurements, TEM images, and the size distributions of particle aggregates are measured by a scanning mobility......The quench cooling of a flame by injection of cold air was studied in a flame reactor for the formation of ZnO particles in a premixed flame with a precursor jet. A rapid temperature drop downstream from the temperature peak is advantageous for the attainment of a large specific surface area....... Computational fluid dynamics simulations were used to design a quench ring with nozzles directed slightly upward and at a small tangential angle from the direct line to the center. This novel design avoids distortion of the flow pattern below the quenching plane and effectively cools the flame immediately above...

  12. CFD Modeling of Sodium-Oxide Deposition in Sodium-Cooled Fast Reactor Compact Heat Exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Tatli, Emre; Ferroni, Paolo; Mazzoccoli, Jason

    2015-09-02

    The possible use of compact heat exchangers (HXs) in sodium-cooled fast reactors (SFR) employing a Brayton cycle is promising due to their high power density and resulting small volume in comparison with conventional shell-and-tube HXs. However, the small diameter of their channels makes them more susceptible to plugging due to Na2O deposition during accident conditions. Although cold traps are designed to reduce oxygen impurity levels in the sodium coolant, their failure, in conjunction with accidental air ingress into the sodium boundary, could result in coolant oxygen levels that are above the saturation limit in the cooler parts of the HX channels. This can result in Na2O crystallization and the formation of solid deposits on cooled channel surfaces, limiting or even blocking coolant flow. The development of analysis tools capable of modeling the formation of these deposits in the presence of sodium flow will allow designers of SFRs to properly size the HX channels so that, in the scenario mentioned above, the reactor operator has sufficient time to detect and react to the affected HX. Until now, analytical methodologies to predict the formation of these deposits have been developed, but never implemented in a high-fidelity computational tool suited to modern reactor design techniques. This paper summarizes the challenges and the current status in the development of a Computational Fluid Dynamics (CFD) methodology to predict deposit formation, with particular emphasis on sensitivity studies on some parameters affecting deposition.

  13. Application of Gamma code coupled with turbomachinery models for high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2008-02-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-ofcoolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of a toxic gas, CO, and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. GAMMA code is being developed to implement turbomachinery models in the power conversion unit (PCU) and ultimately models associated with the hydrogen plant. Some preliminary results will be described in this paper.

  14. Development of GAMMA Code and Evaluation for a Very High Temperature gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Chang H; Lim, H.S.; Kim, E.S.; NO, H.C.

    2007-06-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. This paper will also include what improvements will be made in the Gamma code for the VHTR.

  15. Gas-Cooled Fast Reactor (GFR) FY05 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. Marshall; T. Totemeier; J. Gan; E.E. Feldman; E.A Hoffman; R.F. Kulak; I.U. Therios; C. P. Tzanos; T.Y.C. Wei; L-Y. Cheng; H. Ludewig; J. Jo; R. Nanstad; W. Corwin; V. G. Krishnardula; W. F. Gale; J. W. Fergus; P. Sabharwall; T. Allen

    2005-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radio toxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. Nevertheless, the GFR was chosen as one of only six Generation IV systems to be pursued based on its ability to meet the Generation IV goals in sustainability, economics, safety and reliability, proliferation resistance and physical protection. Current research and development on the Gas-Cooled Fast Reactor (GFR) has focused on the design of safety systems that will remove the decay heat during accident conditions, ion irradiations of candidate ceramic materials, joining studies of oxide dispersion strengthened alloys; and within the Advanced Fuel Cycle Initiative (AFCI) the fabrication of carbide fuels and ceramic fuel matrix materials, development of non-halide precursor low density and high density ceramic coatings, and neutron irradiation of candidate ceramic fuel matrix and metallic materials. The vast majority of this work has focused on the reference design for the GFR: a helium-cooled, direct power conversion system that will operate with on outlet temperature of 850 C at 7 MPa. In addition to the work being performed in the United States, seven international partners under the Generation IV International Forum (GIF) have identified their interest in

  16. Natural Convection and Boiling for Cooling SRP Reactors During Loss of Circulation Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Buckner, M.R.

    2001-06-26

    This study investigated natural convection and boiling as a means of cooling SRP reactors in the event of a loss of circulation accident. These studies show that single phase natural convection cooling of SRP reactors in shutdown conditions with the present piping geometry is probably not feasible.

  17. Air cooled turbine component having an internal filtration system

    Science.gov (United States)

    Beeck, Alexander R [Orlando, FL

    2012-05-15

    A centrifugal particle separator is provided for removing particles such as microscopic dirt or dust particles from the compressed cooling air prior to reaching and cooling the turbine blades or turbine vanes of a turbine engine. The centrifugal particle separator structure has a substantially cylindrical body with an inlet arranged on a periphery of the substantially cylindrical body. Cooling air enters centrifugal particle separator through the separator inlet port having a linear velocity. When the cooling air impinges the substantially cylindrical body, the linear velocity is transformed into a rotational velocity, separating microscopic particles from the cooling air. Microscopic dust particles exit the centrifugal particle separator through a conical outlet and returned to a working medium.

  18. High-temperature gas-cooled reactor safety studies. Progress report for January 1, 1974--June 30, 1975

    Energy Technology Data Exchange (ETDEWEB)

    Cole, T.E.; Sanders, J.P.; Kasten, P.R.

    1977-07-01

    Progress is reported in the following areas: systems and safety analysis; fission product technology; primary coolant technology; seismic and vibration technology; confinement components; primary system materials technology; safety instrumentation; loss of flow accident analysis using HEATUP code; use of coupled-conduction-convection model for core thermal analysis; development of multichannel conduction-convection program HEXEREI; cooling system performance after shutdown; core auxiliary cooling system performance; development of FLODIS code; air ingress into primary systems following DBDA; performance of PCRV thermal barrier cover plates; temperature limits for fuel particle coating failure; tritium distribution and release in HTGR; energy release to PCRV during DBDA; and mathematical models for HTGR reactor safety studies.

  19. Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Greenwood, Michael Scott [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-12-01

    Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

  20. CFD Analysis of the Primary Cooling System for the Small Modular Natural Circulation Lead Cooled Fast Reactor SNRLFR-100

    Directory of Open Access Journals (Sweden)

    Pengcheng Zhao

    2016-01-01

    Full Text Available Small modular reactor (SMR has drawn wide attention in the past decades, and Lead cooled fast reactor (LFR is one of the most promising advanced reactors which are able to meet the safety economic goals of Gen-IV nuclear energy systems. A small modular natural circulation lead cooled fast reactor-100 MWth (SNRLFR-100 is being developed by University of Science and Technology of China (USTC. In the present work, a 3D CFD model, primary heat exchanger model, fuel pin model, and point kinetic model were established based on some reasonable simplifications and assumptions, the steady-state natural circulation characteristics of SNCLFR-100 primary cooling system were discussed and illustrated, and some reasonable suggestions were proposed for the reactor’s thermal-hydraulic and structural design. Moreover, in order to have a first evaluation of the system behavior in accident conditions, an unprotected loss of heat sink (ULOHS transient simulation at beginning of the reactor cycle (BOC has been analyzed and discussed based on the steady-state simulation results. The key temperatures of the reactor core are all under the safety limits at transient state; the reactor has excellent thermal-hydraulic performance.

  1. An alternative solution for heavy liquid metal cooled reactors fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vitale Di Maio, Damiano, E-mail: damiano.vitaledimaio@uniroma1.it [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Cretara, Luca; Giannetti, Fabio [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Peluso, Vincenzo [“ENEA”, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Gandini, Augusto [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy); Manni, Fabio [“SRS Engineering Design S.r.l.”, Vicolo delle Palle 25-25/b, 00186 Rome (Italy); Caruso, Gianfranco [“SAPIENZA” University of Rome – DIAEE, Corso Vittorio Emanuele II, 244, 00186 Rome (Italy)

    2014-10-15

    Highlights: • A new fuel assembly locking system for heavy metal cooled reactor is proposed. • Neutronic, mechanical and thermal-hydraulic evaluations of the system behavior have been performed. • A comparison with other solutions has been presented. - Abstract: In the coming future, the electric energy production from nuclear power plants will be provided by both thermal reactors and fast reactors. In order to have a sustainable energy production through fission reactors, fast reactors should provide an increasing contribution to the total electricity production from nuclear power plants. Fast reactors have to achieve economic and technical targets of Generation IV. Among these reactors, Sodium cooled Fast Reactors (SFRs) and Lead cooled Fast Reactors (LFRs) have the greatest possibility to be developed as industrial power plants within few decades. Both SFRs and LFRs require a great R and D effort to overcome some open issues which affect the present designs (e.g. sodium-water reaction for the SFRs, erosion/corrosion for LFRs, etc.). The present paper is mainly focused on LFR fuel assembly (FA) design: issues linked with the high coolant density of lead or lead–bismuth eutectic cooled reactors have been investigated and an innovative solution for the core mechanical design is here proposed and analyzed. The solution, which foresees cylindrical fuel assemblies and exploits the buoyancy force due to the lead high density, allows to simplify the FAs locking system, to reduce their length and could lead to a more uniform neutron flux distribution.

  2. 14 CFR 29.1109 - Carburetor air cooling.

    Science.gov (United States)

    2010-01-01

    ... 14 Aeronautics and Space 1 2010-01-01 2010-01-01 false Carburetor air cooling. 29.1109 Section 29.1109 Aeronautics and Space FEDERAL AVIATION ADMINISTRATION, DEPARTMENT OF TRANSPORTATION AIRCRAFT... cooling. It must be shown under § 29.1043 that each installation using two-stage superchargers has...

  3. Recent Development of Air-Cooled Bainitic Steels Containing Manganese

    Institute of Scientific and Technical Information of China (English)

    FANG Hong-sheng; YANG Fu-bao; BAI Bing-zhe; YANG Zhi-gang; YIN Jiang

    2005-01-01

    The superiorities of air-cooled bainitic steels were described.A series of air-cooled bainitic steels containing manganese were developed and presented,which include low carbon granular bainitic steels,low carbon grain-boundary allotriomorphic ferrite/granular bainite dual phase steels,medium and medium high carbon bainite/martensite dual phase steels and casting bainitic steels.The development of ultra-low carbon bainitic steels in China was also introduced.

  4. High temperature gas-cooled reactor: gas turbine application study

    Energy Technology Data Exchange (ETDEWEB)

    1980-12-01

    The high-temperature capability of the High-Temperature Gas-Cooled Reactor (HTGR) is a distinguishing characteristic which has long been recognized as significant both within the US and within foreign nuclear energy programs. This high-temperature capability of the HTGR concept leads to increased efficiency in conventional applications and, in addition, makes possible a number of unique applications in both electrical generation and industrial process heat. In particular, coupling the HTGR nuclear heat source to the Brayton (gas turbine) Cycle offers significant potential benefits to operating utilities. This HTGR-GT Application Study documents the effort to evaluate the appropriateness of the HTGR-GT as an HTGR Lead Project. The scope of this effort included evaluation of the HTGR-GT technology, evaluation of potential HTGR-GT markets, assessment of the economics of commercial HTGR-GT plants, and evaluation of the program and expenditures necessary to establish HTGR-GT technology through the completion of the Lead Project.

  5. Creation of Air-Cooled Mn Series Bainitic Steels

    Institute of Scientific and Technical Information of China (English)

    FANG Hong-sheng; FENG Chun; ZHENG Yan-kang; YANG Zhi-gang; BAI Bing-zhe

    2008-01-01

    The development and mechanical performances of new type air-cooled Mn series bainitie steels including granular bainitie steels,FGBA/BG duplex steels,CFB/M duplex steels,medium carbon bainite/martensite steels,cast hainitic steels invented by the authors are summarized.The novel series of bainitie steels are alloyed with Mn,and several series bainitic duplex microstructures can be easily obtained under the condition of air cooling through unique composition design.The invented idea,the principle of alloying design,the strengthening mechanism,and the evolution of the microstructure of new type air-cooled Mn series bainitic steels are presented.Furthermore,the applications in different fields of these Mn series air-cooled bainitic steels with different strength level are also introdueed.It is suggested that the significance of the development of the air-cooled Mn series bainitic steel can be summarized as follows:reducing costs of both raw materials and production;good combination of strength and toughness;self-hardening with high bainitic hardenability by air cooling from hot working without additional quenching-tempering treatment or quenching procedure;large savings in energy resources;and reduced environmental pollution.

  6. 21 CFR 211.46 - Ventilation, air filtration, air heating and cooling.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 4 2010-04-01 2010-04-01 false Ventilation, air filtration, air heating and... Buildings and Facilities § 211.46 Ventilation, air filtration, air heating and cooling. (a) Adequate... holding of a drug product. (c) Air filtration systems, including prefilters and particulate matter...

  7. Ambient air cooling arrangement having a pre-swirler for gas turbine engine blade cooling

    Science.gov (United States)

    Lee, Ching-Pang; Tham, Kok-Mun; Schroeder, Eric; Meeroff, Jamie; Miller, Jr., Samuel R; Marra, John J

    2015-01-06

    A gas turbine engine including: an ambient-air cooling circuit (10) having a cooling channel (26) disposed in a turbine blade (22) and in fluid communication with a source (12) of ambient air: and an pre-swirler (18), the pre-swirler having: an inner shroud (38); an outer shroud (56); and a plurality of guide vanes (42), each spanning from the inner shroud to the outer shroud. Circumferentially adjacent guide vanes (46, 48) define respective nozzles (44) there between. Forces created by a rotation of the turbine blade motivate ambient air through the cooling circuit. The pre-swirler is configured to impart swirl to ambient air drawn through the nozzles and to direct the swirled ambient air toward a base of the turbine blade. The end walls (50, 54) of the pre-swirler may be contoured.

  8. Design and Transient Analysis of Passive Safety Cooling Systems for Advanced Nuclear Reactors

    OpenAIRE

    Galvez, Cristhian

    2011-01-01

    The Pebble Bed Advanced High Temperature Reactor (PB-AHTR) is a pebble fueled, liquid salt cooled, high temperature nuclear reactor design that can be used for electricity generation or other applications requiring the availability of heat at elevated temperatures. A stage in the design evolution of this plant requires the analysis of the plant during a variety of potential transients to understand the primary and safety cooling system response. This study focuses on the performance of the pa...

  9. Can your air-cooled condenser handle the big chill?

    Energy Technology Data Exchange (ETDEWEB)

    Larinoff, M.W.

    1997-01-01

    The two-stage air-cooled condenser (ACC) is less vulnerable to freezeup than the single-stage. But not surprisingly, the greater freeze-protection margin typically costs more. External freeze-protection devices--such as warm-air recirculation systems and airflow control louvers--can mitigate the single-stage design`s vulnerability, although these add to costs, too. Air-cooled condensers have become popular to conserve water, abate plumes, and ease other environmental concerns. This article discusses the critical differences in freeze-protection design features.

  10. Preventing freezing of condensate inside tubes of air cooled condenser

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Jeong A; Hwang, In Hwan; Lee, Dong Hwan [Chonbuk Nat' l Univ., Jeonju (Korea, Republic of); Cho, Young Il [Drexel Univ., Philadelphia (United States)

    2012-08-15

    An air cooled condenser is a device that is used for converting steam into condensate by using ambient air. The air cooled condenser is prone to suffer from a serious explosion when the condensate inside the tubes of a heat exchanger is frozen; in particular, tubes can break during winter. This is primarily due to the structural problem of the tube outlet of an existing conventional air cooled condenser system, which causes the backflow of residual steam and noncondensable gases. To solve the backflow problem in such condensers, such a system was simulated and a new system was designed and evaluated in this study. The experimental results using the simulated condenser showed the occurrence of freezing because of the backflow inside the tube. On the other hand, no backflow and freezing occurred in the advanced new condenser, and efficient heat exchange occurred.

  11. Calculation of Radioactivity and Dose Rate of Activated Corrosion Products in Water-Cooled Fusion Reactor

    Directory of Open Access Journals (Sweden)

    Jingyu Zhang

    2016-01-01

    Full Text Available In water-cooled reactor, the dominant radioactive source term under normal operation is activated corrosion products (ACPs, which have an important impact on reactor inspection and maintenance. A three-node transport model of ACPs was introduced into the new version of ACPs source term code CATE in this paper, which makes CATE capable of theoretically simulating the variation and the distribution of ACPs in a water-cooled reactor and suitable for more operating conditions. For code testing, MIT PWR coolant chemistry loop was simulated, and the calculation results from CATE are close to the experimental results from MIT, which means CATE is available and credible on ACPs analysis of water-cooled reactor. Then ACPs in the blanket cooling loop of water-cooled fusion reactor ITER under construction were analyzed using CATE and the results showed that the major contributors are the short-life nuclides, especially Mn-56. At last a point kernel integration code ARShield was coupled with CATE, and the dose rate around ITER blanket cooling loop was calculated. Results showed that after shutting down the reactor only for 8 days, the dose rate decreased nearly one order of magnitude, which was caused by the rapid decay of the short-life ACPs.

  12. Comparison of sodium and lead-cooled fast reactors regarding reactor physics aspects, severe safety and economical issues

    Energy Technology Data Exchange (ETDEWEB)

    Tucek, Kamil [Joint Research Centre of the European Commission, Institute for Energy, Postbus 2, NL-1755 ZG Petten (Netherlands)]. E-mail: kamil.tucek@jrc.nl; Carlsson, Johan [Joint Research Centre of the European Commission, Institute for Energy, Postbus 2, NL-1755 ZG Petten (Netherlands); Wider, Hartmut [Joint Research Centre of the European Commission, Institute for Energy, Postbus 2, NL-1755 ZG Petten (Netherlands)

    2006-08-15

    A large number of new fast reactors may be needed earlier than foreseen in the Generation IV plans. According to the median forecast of the Special Report on Emission Scenarios commissioned by the Intergovernmental Panel on Climate Control nuclear power will increase by a factor of four by 2050. The drivers for this expected boost are the increasing energy demand in developing countries, energy security, but also climate concerns. However, staying with a once-through cycle will lead to both a substantially increased amount of high-level nuclear waste and an upward pressure on the price of uranium and even concerns about its availability in the coming decades. Therefore, it appears wise to accelerate the development of fast reactors and efficient re-processing technologies. In this paper, two fast reactor systems are discussed-the sodium-cooled fast reactor, which has already been built and can be further improved, and the lead-cooled fast reactor that could be developed relatively soon. An accelerated development of the latter is possible due to the sizeable experience on lead/bismuth eutectic coolant in Russian Alpha-class submarine reactors and the research efforts on accelerator-driven systems in the EU and other countries. First, comparative calculations on critical masses, fissile enrichments and burn-up swings of mid-sized SFRs and LFRs (600 MW{sub e}) are presented. Monte Carlo transport and burn-up codes were used in the analyses. Moreover, Doppler and coolant temperature and axial fuel expansion reactivity coefficients were also evaluated with MCNP and subsequently used in the European Accident Code-2 to calculate reactivity transients and unprotected Loss-of-Flow (ULOF) and Loss-of-Heat Sink (ULOHS) accidents. Further, ULOFs as well as decay heat removal (protected Total Loss-of-Power, TLOP) were calculated with the STAR-CD CFD code for both systems. We show that LFRs and SFRs can be used both as burners and as self-breeders, homogeneously incinerating

  13. Closed Fuel Cycle and Minor Actinide Multirecycling in a Gas-Cooled Fast Reactor

    NARCIS (Netherlands)

    Van Rooijen, W.F.G.; Kloosterman, J.L.

    2009-01-01

    The Generation IV International Forum has identified the Gas-Cooled Fast Reactor (GCFR) as one of the reactor concepts for future deployment. The GCFR targets sustainability, which is achieved by the use of a closed nuclear fuel cycle where only fission products are discharged to a repository; all H

  14. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-01-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

  15. High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

    Energy Technology Data Exchange (ETDEWEB)

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

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

    OpenAIRE

    Sungjoo Lee; Byungun Yoon; Juneseuk Shin

    2016-01-01

    We propose a stepwise method of selecting appropriate indicators to measure effects of a specific nuclear energy option on sustainable development and energy security, and also to compare an energy option with another. Focusing on the sodium-cooled fast reactor, one of the highlighted Generation IV reactors, we measure and compare its effects with the standard pressurized water reactor-based nuclear power, and then with coal power. Collecting 36 indicators, five experts select seven key indic...

  17. Challenges and Innovative Technologies On Fuel Handling Systems for Future Sodium-Cooled Fast Reactors

    OpenAIRE

    Chassignet, Mathieu; Dumas, Sebastien; Penigot, Christophe; Prele, Gerard; Capitaine, Alain; Rodriguez, Gilles; Sanseigne, Emmanuel; Beauchamp, Francois

    2011-01-01

    International audience; The reactor refuelling system provides the means of transporting, storing, and handling reactor core subassemblies. The system consists of the facilities and equipment needed to accomplish the scheduled refuelling operations. The choice of a FHS impacts directly on the general design of the reactor vessel (primary vessel, storage, and final cooling before going to reprocessing), its construction cost, and its availability factor. Fuel handling design must take into acc...

  18. Air cooling using a matrix of ceramic tubes

    Science.gov (United States)

    Ali, M.; Zeitoun, O.; Al-Ansary, H.; Nuhait, A.

    2012-05-01

    An experimental study is conducted to cool the outdoor air using a humidification technique. A wind tunnel was built with a matrix of ceramic tube test section. An outdoor air passes over the ceramic tube matrix (cross flow) where water passing through the ceramic tubes. Air temperatures and relative humidity are measured before and after the test section for several air and water speeds. Air speed is measured at different locations along the centerline of the cross section. Results show that the ambient temperature drops by about 10 °C when the relative humidity increases from 2% to 5.4%.

  19. Topical report: Natural convection shutdown heat removal test facility (NSTF) evaluation for generating additional reactor cavity cooling system (RCCS) data.

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, M. T.; Kilsdonk, D. J.; Tzanos, C.P.; Lomperski, S.; Aeschlimann, R.W.; Pointer, D.; Nuclear Engineering Division

    2005-09-01

    As part of the Department of Energy (DOE) Generation IV roadmapping activity, the Very High Temperature gas cooled Reactor (VHTR) has been selected as the principal concept for hydrogen production and other process-heat applications such as district heating and potable water production. On this basis, the DOE has selected the VHTR for additional R&D with the ultimate goal of demonstrating emission-free electricity and hydrogen production with this advanced reactor concept. One of the key passive safety features of the VHTR is the potential for decay heat removal by natural circulation of air in a Reactor Cavity Cooling System (RCCS). The air-cooled RCCS concept is notably similar to the Reactor Vessel Auxiliary Cooling System (RVACS) that was developed for the General Electric PRISM sodium-cooled fast reactor. As part of the DOE R&D program that supported the development of this fast reactor concept, the Natural Convection Shutdown Heat Removal Test Facility (NSTF) was developed at ANL to provide proof-of-concept data for the RVACS under prototypic natural convection flow, temperature, and heat flux conditions. Due to the similarity between RVACS and the RCCS, current VHTR R&D plans call for the utilization of the NSTF to provide RCCS model development and validation data, in addition to supporting design validation and optimization activities. Both air-cooled and water-cooled RCCS designs are to be included. In support of this effort, ANL has been tasked with the development of an engineering plan for mechanical and instrumentation modifications to NSTF to ensure that sufficiently detailed temperature, heat flux, velocity and turbulence profiles are obtained to adequately qualify the codes under the expected range of air-cooled RCCS flow conditions. Next year, similar work will be carried out for the alternative option of a water-cooled RCCS design. Analysis activities carried out in support of this experiment planning task have shown that: (a) in the RCCS, strong

  20. Conceptual design of a passive moderator cooling system for a pressure tube type natural circulation boiling water cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Pal, Eshita, E-mail: eshi.pal@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Nayak, Arun K.; Vijayan, Pallipattu K. [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

    2015-09-15

    Highlights: • Passive moderator cooling system is designed to cool moderator passively during SBO. • PMCS is a system of two natural circulation loops, coupled via a heat exchanger. • RELAP5 analyses show that PMCS maintains moderator within safe limits for 7 days. - Abstract: The recent Fukushima accident has raised strong concern and apprehensions about the safety of reactors in case of a prolonged Station Black Out (SBO) continuing for several days. In view of this, a detailed study was performed simulating this condition in Advanced Heavy Water Reactor. In this study, a novel concept of moderator cooling by passive means has been introduced in the reactor design. The Passive Moderator Cooling System (PMCS) consists of a shell and tube heat exchanger designed to remove 2 MW heat from the moderator inside Calandria. The heat exchanger is located at a suitable elevation from the Calandria of the reactor, such that the hot moderator rises due to buoyancy into the heat exchanger and upon cooling from shell side water returns to Calandria forming a natural circulation loop. The shell side of the heat exchanger is also a natural circulation loop connected to an overhead large water reservoir, namely the GDWP. The objective of the PMCS is to remove the heat from the moderator in case of an SBO and maintaining its temperature below the permissible safe limit (100 °C) for at least 7 days. The paper first describes the concept of the PMCS. The concept has been assessed considering a prolonged SBO for at least 7 days, through an integrated analysis performed using the code RELAP5/MOD3.2 considering all the major components of the reactor. The analysis shows that the PMCS is able to maintain the moderator temperature below boiling conditions for 7 days.

  1. Studies on advanced water-cooled reactors beyond generation Ⅲ for power generation

    Institute of Scientific and Technical Information of China (English)

    CHENG Xu

    2007-01-01

    China's ambitious nuclear power program motivates the country's nuclear community to develop advanced reactor concepts beyond generation Ⅲ to ensure a long-term, stable, and sustainable development of nuclear power. The paper discusses some main criteria for the selection of future water-cooled reactors by considering the specific Chinese situation. Based on the suggested selection criteria, two new types of water-cooled reactors are recommended for future Chinese nuclear power generation. The high conversion pressurized water reactor utilizes the present PWR technology to a large extent. With a conversion ratio of about 0.95, the fuel utilization is increased about 5 times. This significantly improves the sustainability of fuel resources. The supercritical water-cooled reactor has favorable features in economics,sustainability and technology availability. It is a logical extension of the generation Ⅲ PWR technology in China.The status of international R&D work is reviewed. A new supercritieal water-cooled reactor (SCWR) core structure (the mixed reactor core) and a new fuel assembly design (two-rows FA) are proposed. The preliminary analysis using a coupled neutron-physics/thermal-hydranlics method is carded out. It shows good feasibility for the new design proposal.

  2. Thermal computations for electronics conductive, radiative, and convective air cooling

    CERN Document Server

    Ellison, Gordon

    2010-01-01

    IntroductionPrimary mechanisms of heat flowConductionApplication example: Silicon chip resistance calculationConvectionApplication example: Chassis panel cooled by natural convectionRadiationApplication example: Chassis panel cooled only by radiation 7Illustrative example: Simple thermal network model for a heat sinked power transistorIllustrative example: Thermal network circuit for a printed circuit boardCompact component modelsIllustrative example: Pressure and thermal circuits for a forced air cooled enclosureIllustrative example: A single chip package on a printed circuit board-the proble

  3. Use of a temperature-initiated passive cooling system (TIPACS) for the modular high-temperature gas-cooled reactor cavity cooling system (RCCS)

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Conklin, J.; Reich, W.J.

    1994-04-01

    A new type of passive cooling system has been invented (Forsberg 1993): the Temperature-Initiated Passive Cooling System (TIPACS). The characteristics of the TIPACS potentially match requirements for an improved reactor-cavity-cooling system (RCCS) for the modular high-temperature gas-cooled reactor (MHTGR). This report is an initial evaluation of the TIPACS for the MHTGR with a Rankines (steam) power conversion cycle. Limited evaluations were made of applying the TIPACS to MHTGRs with reactor pressure vessel temperatures up to 450 C. These temperatures may occur in designs of Brayton cycle (gas turbine) and process heat MHTGRs. The report is structured as follows. Section 2 describes the containment cooling issues associated with the MHTGR and the requirements for such a cooling system. Section 3 describes TIPACS in nonmathematical terms. Section 4 describes TIPACS`s heat-removal capabilities. Section 5 analyzes the operation of the temperature-control mechanism that determines under what conditions the TIPACS rejects heat to the environment. Section 6 addresses other design and operational issues. Section 7 identifies uncertainties, and Section 8 provides conclusions. The appendixes provide the detailed data and models used in the analysis.

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

    Science.gov (United States)

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

    2014-08-01

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

  5. Cool Colored Roofs to Save Energy and Improve Air Quality

    Energy Technology Data Exchange (ETDEWEB)

    Akbari, Hashem; Levinson, Ronnen; Miller, William; Berdahl, Paul

    2005-08-23

    Urban areas tend to have higher air temperatures than their rural surroundings as a result of gradual surface modifications that include replacing the natural vegetation with buildings and roads. The term ''Urban Heat Island'' describes this phenomenon. The surfaces of buildings and pavements absorb solar radiation and become extremely hot, which in turn warm the surrounding air. Cities that have been ''paved over'' do not receive the benefit of the natural cooling effect of vegetation. As the air temperature rises, so does the demand for air-conditioning (a/c). This leads to higher emissions from power plants, as well as increased smog formation as a result of warmer temperatures. In the United States, we have found that this increase in air temperature is responsible for 5-10% of urban peak electric demand for a/c use, and as much as 20% of population-weighted smog concentrations in urban areas. Simple ways to cool the cities are the use of reflective surfaces (rooftops and pavements) and planting of urban vegetation. On a large scale, the evapotranspiration from vegetation and increased reflection of incoming solar radiation by reflective surfaces will cool a community a few degrees in the summer. As an example, computer simulations for Los Angeles, CA show that resurfacing about two-third of the pavements and rooftops with reflective surfaces and planting three trees per house can cool down LA by an average of 2-3K. This reduction in air temperature will reduce urban smog exposure in the LA basin by roughly the same amount as removing the basin entire onroad vehicle exhaust. Heat island mitigation is an effective air pollution control strategy, more than paying for itself in cooling energy cost savings. We estimate that the cooling energy savings in U.S. from cool surfaces and shade trees, when fully implemented, is about $5 billion per year (about $100 per air-conditioned house).

  6. Safety Analysis of Pb-208 Cooled 800 MWt Modified CANDLE Reactors

    Science.gov (United States)

    Su'ud, Zaki; Widiawati, Nina; Sekimoto, H.; Artoto, A.

    2017-01-01

    Safely analysis of 800MWt Pb-208 cooled fast reactors with natural Uranium as fuel cycle input employing axial-radial combined Modiified CANDLE burnup scheme has been performed. The analysis of unprotected loss of flow(ULOF) and unprotected rod run-out transient overpower (UTOP) are discussed. Some simulations for 800 MWt Pb-208 cooled fast reactors has been performed and the results show that the reactor can anticipate complete pumping failure inherently by reducing power through reactivity feedback and remove the rest of heat through natural circulations. Compared to the Pb-nat cooled long life Fast Reactors, Pb-208 cooled reactors have smaller Doppler but higher coolant density reactivity coefficient. In the UTOP accident case the analysis has been performed against external reactivity up to 0.003dk/k. And for ULOHS case it is assumed that the secondary cooling system has broken. During all accident the cladding temperature is the most critical. Especially for the case of UTOP accident. In addition the steam generator design has also consider excess power which may reach 50% extra during severe UTOP case..

  7. A review of gas-cooled reactor concepts for SDI (Strategic Defense Initiative) applications

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, A.C.

    1989-08-01

    We have completed a review of multimegawatt gas-cooled reactor concepts proposed for SDI applications. Our study concluded that the principal reason for considering gas-cooled reactors for burst-mode operation was the potential for significant system mass savings over closed-cycle systems if open-cycle gas-cooled operation (effluent exhausted to space) is acceptable. The principal reason for considering gas-cooled reactors for steady-state operation is that they may represent a lower technology risk than other approaches. In the review, nine gas-cooled reactor concepts were compared to identify the most promising. For burst-mode operation, the NERVA (Nuclear Engine for Rocket Vehicle Application) derivative reactor concept emerged as a strong first choice since its performance exceeds the anticipated operational requirements and the technology has been demonstrated and is retrievable. Although the NERVA derivative concepts were determined to be the lead candidates for the Multimegawatt Steady-State (MMWSS) mode as well, their lead over the other candidates is not as great as for the burst mode. 90 refs., 2 figs., 10 tabs.

  8. Bioaerosol deposition on an air-conditioning cooling coil

    Science.gov (United States)

    Wu, Yan; Chen, Ailu; Luhung, Irvan; Gall, Elliott T.; Cao, Qingliang; Chang, Victor Wei-Chung; Nazaroff, William W.

    2016-11-01

    This study is concerned with the role of a fin-and-tube heat exchanger in modifying microbial indoor air quality. Specifically, depositional losses of ambient bioaerosols and particles onto dry (not cooled) and wet (cool) coil surfaces were measured for different airspeeds passing through the test coil. Total, bacterial and fungal DNA concentrations in condensate water produced by a wet coil were also quantified by means of fluorescent dsDNA-binding dye and qPCR assays. Results revealed that the deposition of bioaerosols and total particles is substantial on coil surfaces, especially when wet and cool. The average deposition fraction was 0.14 for total DNA, 0.18 for bacterial DNA and 0.22 for fungal DNA on the dry coil, increasing to 0.51 for total DNA, 0.50 for bacterial DNA and 0.68 for fungal DNA on the wet coil. Overall, as expected, deposition fractions increased with increasing particle size and increasing airspeed. Deposited DNA was removed from the cooling coil surfaces through the flow of condensing water at a rate comparable to the rate of direct deposition from air. A downward trend of bacterial and fungal DNA measured in condensate water over time provides suggestive evidence of biological growth on heat exchangers during nonoperational times of a ventilation system. This investigation provides new information about bioaerosol deposition onto a conventional fin-and-tube cooling coil, a potentially important factor influencing indoor exposure to microbial aerosols in air-conditioned buildings.

  9. Thermal Hydraulics of the Very High Temperature Gas Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh; Eung Kim; Richard Schultz; Mike Patterson; Davie Petti

    2009-10-01

    The U.S Department of Energy (DOE) is conducting research on the Very High Temperature Reactor (VHTR) design concept for the Next Generation Nuclear Plant (NGNP) Project. The reactor design will be a graphite moderated, thermal neutron spectrum reactor that will produce electricity and hydrogen in a highly efficient manner. The NGNP reactor core will be either a prismatic graphite block type core or a pebble bed core. The NGNP will use very high-burnup, low-enriched uranium, TRISO-coated fuel, and have a projected plant design service life of 60 years. The VHTR concept is considered to be the nearest-term reactor design that has the capability to efficiently produce hydrogen. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during reactor core-accidents. The objectives of the NGNP Project are to: Demonstrate a full-scale prototype VHTR that is commercially licensed by the U.S. Nuclear Regulatory Commission, and Demonstrate safe and economical nuclear-assisted production of hydrogen and electricity. The DOE laboratories, led by the INL, perform research and development (R&D) that will be critical to the success of the NGNP, primarily in the areas of: • High temperature gas reactor fuels behavior • High temperature materials qualification • Design methods development and validation • Hydrogen production technologies • Energy conversion. This paper presents current R&D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs.

  10. Study on forced air convection cooling for electronic assemblies

    Institute of Scientific and Technical Information of China (English)

    2008-01-01

    The slotted fin concept was employed to improve the air cooling performance of plate-fin in heat sinks.Numerical simulations of laminar heat transfer and flow pressure drop were conducted for the integral plate fin,discrete plate fin and discrete slotted fin heat sinks.It is found that the performance of the discrete plate fin is better than that of the integral continuum plate fin and the performance of slotted fin is better than that of the discrete plate fin at the same pumping power of the fan.A new type of heat sink characterized by discrete and slotted fin surfaces with thinner fins and smaller spaces between fins is then proposed.Preliminary computation shows that this type of heat sink may be useful for the next generation of higher thermal load CPUs.The limit of cooling capacity for air-cooling techniques was also addressed.

  11. Preliminary Study of Lead-Oxide Cooled Fast Reactor with Natural Uranium as an Input Fuel with Reactor Shuffling Strategy

    Science.gov (United States)

    Mahmudah, Rida SN; Su’ud, Zaki

    2017-01-01

    A preliminary study of lead-oxide cooled fast reactor with natural uranium as an input fuel using reactor shuffling strategy has been conducted. In this study, reactor core is divided into four zone with the same volume, each zone use different uranium enrichment. The enrichment number is estimated so that in the end of reactor’s operation, we only need to add natural uranium as the fresh input fuel. This study used UN-PuN as the fuel and lead oxide as the coolant. Several parameter studies have been conducted to determine the most suitable input condition. It is confirmed in this study that with fuel : cladding : coolant ratio of 53 : 10 : 37, and uranium enrichment in the first to the fourth zone of 0%, 6.25%, 7.5% and 8%, respectively, the reactor can operate as long as 20 years of operation with terminal k-eff of 1.0004.

  12. Air cooling effect of fins on a Honda shine bike

    Directory of Open Access Journals (Sweden)

    Padhiyar Abhesinh J

    2015-05-01

    Full Text Available The main of aim of this work is to study various researches done in past to improve heat transfer rate of cooling fins by changing cylinder block fin geometry. Low rate of heat transfer through cooling fins is the main problem in this type of cooling. So efficiency of the engine is increase by increase the heat transfer. Examples of direct air cooling in modern automobiles are rare. The most common example is the commercials Automobile bike like a Honda Shine, Bajaj bike, Honda splendor etc. It is conclude about shape try to this fins is more effectively heat transfer in Honda shine bike compare to existing fins. After FEA Analysis it checking on fin whether efficiency of heat transfer increases or not. This work validation with Experimental and Mathematical.

  13. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    Science.gov (United States)

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  14. Gas-Cooled Thorium Reactor with Fuel Block of the Unified Design

    Directory of Open Access Journals (Sweden)

    Igor Shamanin

    2015-01-01

    Full Text Available Scientific researches of new technological platform realization carried out in Russia are based on ideas of nuclear fuel breeding in closed fuel cycle and physical principles of fast neutron reactors. Innovative projects of low-power reactor systems correspond to the new technological platform. High-temperature gas-cooled thorium reactors with good transportability properties, small installation time, and operation without overloading for a long time are considered perspective. Such small modular reactor systems at good commercial, competitive level are capable of creating the basis of the regional power industry of the Russian Federation. The analysis of information about application of thorium as fuel in reactor systems and its perspective use is presented in the work. The results of the first stage of neutron-physical researches of a 3D model of the high-temperature gas-cooled thorium reactor based on the fuel block of the unified design are given. The calculation 3D model for the program code of MCU-5 series was developed. According to the comparison results of neutron-physical characteristics, several optimum reactor core compositions were chosen. The results of calculations of the reactivity margins, neutron flux distribution, and power density in the reactor core for the chosen core compositions are presented in the work.

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

    Directory of Open Access Journals (Sweden)

    Sungjoo Lee

    2016-09-01

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

  16. Cryogenic Cooling System for 5 kA, 200 μH Class HTS DC Reactor

    Science.gov (United States)

    Park, Heecheol; Kim, Seokho; Kim, Kwangmin; Park, Minwon; Park, Taejun; Kim, A.-rong; Lee, Sangjin

    DC reactors, made by aluminum busbar, are used to stabilize the arc of an electric furnace. In the conventional arc furnace, the transport current is several tens of kilo-amperes and enormous resistive loss is generated. To reduce the resistive loss at the DC reactor, a HTS DC reactor can be considered. It can dramatically improve the electric efficiency as well as reduce the installation space. Similar with other superconducting devices, the HTS DC reactor requires current leads from a power source in room temperature to the HTS coil in cryogenic environment. The heat loss at the metal current leads can be minimized through optimization process considering the geometry and the transport current. However, the transport current of the HTS DC reactor for the arc furnace is much larger than most of HTS magnets and the enormous heat penetration through the current lead should be effectively removed to keep the temperature around 70∼77 K. Current leads are cooled down by circulation of liquid nitrogen from the cooling system with a stirling cryocooler. The operating temperature of HTS coil is 30∼40 K and circulation of gaseous helium is used to remove the heat generation at the HTS coil. Gaseous helium is transported through the cryogenic helium blower and a single stage GM cryocooler. This paper describes design and experimental results on the cooling system for current leads and the HTS coil of 5 kA, 200 μH class DC reactor as a prototype. The results are used to verify the design values of the cooling systems and it will be applied to the design of scale-up cooling system for 50 kA, 200 μH class DC reactor.

  17. Circulating and plateout activity program for gas-cooled reactors with arbitrary radioactive chains

    Energy Technology Data Exchange (ETDEWEB)

    Apperson, C.E. Jr.

    1978-03-01

    A time-dependent method for estimating the fuel body, circulating, plateout, and filter inventory of a high temperature gas-cooled reactor (HTGR) during normal operation is discussed. The primary coolant model accounts for the source, buildup, decay, and cleanup of isotopes that are gas borne inside the prestressed concrete reactor vessel (PCRV). This method has been implemented in the SUVIUS computer program that is described in detail.

  18. Device for cooling the main vessel of a fast fission nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Debru, M.

    1984-10-16

    The annular space delimited by the main vessel and an internal shell is in communication with the zone of the reactor vessel, in which the cold primary liquid is located. The annular space delimited by the shell and by an internal shell is in communication with the lower part of the core via tubes. Thus, the cold primary liquid is injected into the space where it circulates from bottom to top, and flows into the space, where it circulates from top to bottom while at the same time cooling the main vessel. The invention applies, in particular, to fast fission nuclear reactors cooled by liquid sodium.

  19. Waste heat recovery using looped heat pipes for air cooling

    Energy Technology Data Exchange (ETDEWEB)

    Lamfon, N.J.; Akyurt, M.; Najjar, Y.S.H. (King Abdulaziz Univ., Jeddah (Saudi Arabia). Mechanical Engineering Dept.)

    1994-07-01

    A scheme is described for the recovery of waste heat from stacks of gas turbine engines and the utilization of recovered energy for the cooling of ambient air. Relationships are summarized for the modeling of components of the cooling system. Samples are presented from performance data that is predicted by the model. Effect of size and design of system components, as well as operational variables on system performance, are discussed. It is concluded that the single most significant variable in the design of the looped heat-pipe recovery and utilization system is the geometry of the exhaust pipe of the gas turbine engine. (author)

  20. Optimal air-supply mode of hybrid system with radiant cooling and dedicated outdoor air

    Institute of Scientific and Technical Information of China (English)

    丁研; 田喆; 朱能

    2015-01-01

    The hybrid system with radiant cooling and dedicated outdoor air not only possesses high energy efficiency, but also creates a healthy and comfortable indoor environment. Indoor air quality will be improved by the dedicated outdoor air system (DOAS) and indoor thermal comfort can be enhanced by the radiant cooling system (RCS). The optimal air-supply mode of the hybrid system and the corresponding design approach were investigated. A full-scale experimental chamber with various air outlets and the ceiling radiant cooling panels (CRCP) was designed and established. The performances of different air-supply modes along with CRCPs were analyzed by multi-index evaluations. Preliminary investigations were also conducted on the humidity stratification and the control effect of different airflow modes to prevent condensation on CRCP. The overhead supply air is recommended as the best combination mode for the hybrid system after comprehensive comparison of the experiment results. The optimal proportion of CRCP accounting for the total cooling capacities in accord with specific cooling loads is found, which may provide valuable reference for the design and operation of the hybrid system.

  1. Gas Cooled, Natural Uranium, D20 Moderated Power Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dahlberg, R.C.; Beasley, E.G.; DeBoer, T.K.; Evans, T.C.; Molino, D.F.; Rothwell, W.S.; Slivka, W.R.

    1956-08-01

    The attractiveness of a helium cooled, heavy water moderated, natural uranium central station power plant has been investigated. A fuel element has been devised which allows the D20 to be kept at a low pressure while the exit gas temperature is high. A preliminary cost analysis indicates that, using currently available materials, competitive nuclear power in foreign countries is possible.

  2. Summarized compatibility review of reactor materials for CO2-cooled graphite-moderated reactors

    Energy Technology Data Exchange (ETDEWEB)

    Seddon, B.J.

    1964-09-23

    This report, which is a revised edition of TRG-Report-267, summarises an internal document and collates information on the compatibility of a range of materials used in CO{sub 2}-cooled graphite-moderated reactors. Information is presented in the form of six tables based on compatibilities of materials with carbon dioxide, beryllium, Magnox, magnesium, uranium and compatibilities of pairs of other relevant materials.

  3. Development of safety analysis codes and experimental validation for a very high temperature gas-cooled reactor Final report

    Energy Technology Data Exchange (ETDEWEB)

    Chang Oh

    2006-03-01

    The very high-temperature gas-cooled reactor (VHTR) is envisioned as a single- or dual-purpose reactor for electricity and hydrogen generation. The concept has average coolant temperatures above 9000C and operational fuel temperatures above 12500C. The concept provides the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperature to support process heat applications, such as coal gasification, desalination or cogenerative processes, the VHTR’s higher temperatures allow broader applications, including thermochemical hydrogen production. However, the very high temperatures of this reactor concept can be detrimental to safety if a loss-of-coolant accident (LOCA) occurs. Following the loss of coolant through the break and coolant depressurization, air will enter the core through the break by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structure and fuel. The oxidation will accelerate heatup of the reactor core and the release of toxic gasses (CO and CO2) and fission products. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. Prior to the start of this Korean/United States collaboration, no computer codes were available that had been sufficiently developed and validated to reliably simulate a LOCA in the VHTR. Therefore, we have worked for the past three years on developing and validating advanced computational methods for simulating LOCAs in a VHTR. Research Objectives As described above, a pipe break may lead to significant fuel damage and fission product release in the VHTR. The objectives of this Korean/United States collaboration were to develop and validate advanced computational methods for VHTR safety analysis. The methods that have been developed are now

  4. Heat pipe cooled reactors for multi-kilowatt space power supplies

    Energy Technology Data Exchange (ETDEWEB)

    Ranken, W.A.; Houts, M.G.

    1995-01-01

    Three nuclear reactor space power system designs are described that demonstrate how the use of high temperature heat pipes for reactor heat transport, combined with direct conversion of heat to electricity, can result in eliminating pumped heat transport loops for both primary reactor cooling and heat rejection. The result is a significant reduction in system complexity that leads to very low mass systems with high reliability, especially in the power range of 1 to 20 kWe. In addition to removing heat exchangers, electromagnetic pumps, and coolant expansion chambers, the heat pipe/direct conversion combination provides such capabilities as startup from the frozen state, automatic rejection of reactor decay heat in the event of emergency or accidental reactor shutdown, and the elimination of single point failures in the reactor cooling system. The power system designs described include a thermoelectric system that can produce 1 to 2 kWe, a bimodal modification of this system to increase its power level to 5 kWe and incorporate high temperature hydrogen propulsion capability, and a moderated thermionic reactor concept with 5 to 20 kWe power output that is based on beryllium modules that thermally couple cylindrical thermionic fuel elements (TFEs) to radiator heat pipes.

  5. Heat pipe cooled reactors for multi-kilowatt space power supplies

    Science.gov (United States)

    Ranken, W. A.; Houts, M. G.

    Three nuclear reactor space power system designs are described that demonstrate how the use of high temperature heat pipes for reactor heat transport, combined with direct conversion of heat to electricity, can result in eliminating pumped heat transport loops for both primary reactor cooling and heat rejection. The result is a significant reduction in system complexity that leads to very low mass systems with high reliability, especially in the power range of 1 to 20 kWe. In addition to removing heat exchangers, electromagnetic pumps, and coolant expansion chambers, the heat pipe/direct conversion combination provides such capabilities as startup from the frozen state, automatic rejection of reactor decay heat in the event of emergency or accidental reactor shutdown, and the elimination of single point failures in the reactor cooling system. The power system designs described include a thermoelectric system that can produce 1 to 2 kWe, a bimodal modification of this system to increase its power level to 5 kWe and incorporate high temperature hydrogen propulsion capability, and a moderated thermionic reactor concept with 5 to 20 kWe power output that is based on beryllium modules that thermally couple cylindrical thermionic fuel elements (TFE's) to radiator heat pipes.

  6. 78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

    Science.gov (United States)

    2013-10-25

    ... COMMISSION Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors AGENCY... Systems for Light-Water-Cooled Nuclear Power Reactors,'' in which the NRC made editorial corrections and... analysis for liquid and gaseous radwaste system components for light water nuclear power...

  7. Experimental study of the decrease in the temperature of an air/water-cooled turbine blade

    Science.gov (United States)

    Ryzhov, A. A.; Sereda, A. V.; Shaiakberov, V. F.; Iskakov, K. M.; Shatalov, Iu. S.

    Results of the full-scale testing of an air/water-cooled deflector-type turbine blade are reported. Data on the decrease in the temperature of the cooling air and of the blade are presented and compared with the calculated values. An analysis of the results indicates that the use of air/water cooling makes it possible to significantly reduce the temperature of the cooling air and of the blade with practically no increase in the engine weight and dimensions.

  8. Experimental and numerical study of open-air active cooling

    Science.gov (United States)

    Al-Fifi, Salman Amsari

    The topic of my thesis is Experimental and Numerical Study of Open Air Active Cooling. The present research is intended to investigate experimentally and Numerically the effectiveness of cooling large open areas like stadiums, shopping malls, national gardens, amusement parks, zoos, transportation facilities and government facilities or even in buildings outdoor gardens and patios. Our cooling systems are simple cooling fans with different diameters and a mist system. This type of cooling systems has been chosen among the others to guarantee less energy consumption, which will make it the most favorable and applicable for cooling such places mentioned above. In the experiments, the main focus is to study the temperature domain as a function of different fan diameters aerodynamically similar in different heights till we come up with an empirical relationship that can determine the temperature domain for different fan diameters and for different heights of these fans. The experimental part has two stages. The first stage is devoted to investigate the maximum range of airspeed and profile for three different fan diameters and for different heights without mist, while the second stage is devoted to investigate the maximum range of temperature and profile for the three different diameter fans and for different heights with mist. The computational study is devoted to built an experimentally verified mathematical model to be used in the design and optimization of water mist cooling systems, and to compare the mathematical results to the experimental results and to get an insight of how to apply such evaporative mist cooling for different places for different conditions. In this study, numerical solution is presented based on experimental conditions, such dry bulb temperature, wet bulb temperature, relative humidity, operating pressure and fan airspeed. In the computational study, all experimental conditions are kept the same for the three fans except the fan airspeed

  9. Report of study 7.3: cooling and air conditioning

    Energy Technology Data Exchange (ETDEWEB)

    Russo, F.

    2000-07-01

    This report describes the results of the study carried out by the study group 7.3 in the triennium 1997-2000. The study was focused on industrial refrigeration and air conditioning for the large building utilising natural gas. The goal of this study, carried out in collaboration of the members of study group 7.3, was to analyse the markets of industrial refrigeration and air conditioning for large buildings to identify possibilities to increase the natural gas share in these sectors. The available technology in the two sectors of the market are described in a single section, i.e. the 'State of the art of the technology'. In this section, technical characteristics, applications, performances, new developments and others topics are discussed for absorbers, gas engines, gas turbines and fuel cells. In the 'Industrial Refrigeration' section an analysis of the present global market for the industrial sector is presented. Economics, advantages and barriers to gas units compared with the electrical units are discussed. Information on existing industrial plants, possible application options and new technology developments are described as well. The 'Air conditioning for the large building' section deals with offices, hotels, commercial buildings, hospitals and shopping centres with a cooling capacity of 350 kW or higher. It appears that the use of natural gas for cooling of large buildings has been increasing during the last decade, thanks to the greater availability of natural gas and the development of new technologies. A marketing survey of gas air-conditioning was carried out in cooperation with a group of Intergas Marketing. Based on the survey, the report describes the market position of natural gas relative to electricity. It provides the strategic prospects for further developing natural gas as a competitive option for air-conditioning of large buildings using a combination of state-of-the-art technologies. It is important to highlight

  10. MR-6 type fuel elements cooling in natural convection conditions after the reactor shut down

    Energy Technology Data Exchange (ETDEWEB)

    Pytel, K.; Bykowski, W.; Moldysz, A. [Institute of Atomic Energy, Otwock Swierk (Poland)

    2002-07-01

    Natural cooling conditions of the nuclear fuel in the channel type reactor after its shut down are commonly determined with relatively high uncertainty. This is not only to he lack of adequate measurements of thermal parameters i.e. the residual power generation, the coolant flow and temperatures, but also due to indeterminate model of convection mechanism. The numerical simulation of natural convection in multitube fuel assembly in the fuel channel leads to various convection modes including evidently chaotic behaviour. To determine the real cooling conditions in the MARIA research reactor a series of experiments has been performed with fuel assembly equipped with a set of thermocouples. After some forced cooling period (the shortest was half an hour after the reactor shut down) the reactor was left with the only natural convection. Two completely different cooling modes have been observed. The MARIA core consists of series of individual fuel channel and so called bypasses, maintaining the hydraulic properties of the fuel channel, connected in parallel. Initially, the convection cells were established trough few so-called bypasses providing a very effective mode of cooling. In this mode the flow charts were identical to those existing in forced cooling mode. After certain period the system switched on the second cooling mode with natural circulation within the individual fuel cells. Higher temperatures and temperature fluctuations were characteristic for this mode approaching 30 deg in amplitude. In almost all the cases the system was switching few times between modes, but eventually remained in the second mode. The switching times were not regular and the process has a chaotic behaviour. (author)

  11. Emperor penguin body surfaces cool below air temperature.

    Science.gov (United States)

    McCafferty, D J; Gilbert, C; Thierry, A-M; Currie, J; Le Maho, Y; Ancel, A

    2013-06-23

    Emperor penguins Aptenodytes forsteri are able to survive the harsh Antarctic climate because of specialized anatomical, physiological and behavioural adaptations for minimizing heat loss. Heat transfer theory predicts that metabolic heat loss in this species will mostly depend on radiative and convective cooling. To examine this, thermal imaging of emperor penguins was undertaken at the breeding colony of Pointe Géologie in Terre Adélie (66°40' S 140° 01' E), Antarctica in June 2008. During clear sky conditions, most outer surfaces of the body were colder than surrounding sub-zero air owing to radiative cooling. In these conditions, the feather surface will paradoxically gain heat by convection from surrounding air. However, owing to the low thermal conductivity of plumage any heat transfer to the skin surface will be negligible. Future thermal imaging studies are likely to yield further insights into the adaptations of this species to the Antarctic climate.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-15

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

  13. Zirconium carbide coating for corium experiments related to water-cooled and sodium-cooled reactors

    Science.gov (United States)

    Plevacova, K.; Journeau, C.; Piluso, P.; Zhdanov, V.; Baklanov, V.; Poirier, J.

    2011-07-01

    Since the TMI and Chernobyl accidents the risk of nuclear severe accident is intensively studied for existing and future reactors. In case of a core melt-down accident in a nuclear reactor, a complex melt, called corium, forms. To be able to perform experiments with prototypic corium materials at high temperature, a coating which resists to different corium melts related to Generation I and II Water Reactors and Generation IV sodium fast reactor was researched in our experimental platforms both in IAE NNC in Kazakhstan and in CEA in France. Zirconium carbide was selected as protective coating for graphite crucibles used in our induction furnaces: VCG-135 and VITI. The method of coating application, called reactive wetting, was developed. Zirconium carbide revealed to resist well to the (U x, Zr y)O 2-z water reactor corium. It has also the advantage not to bring new elements to this chemical system. The coating was then tested with sodium fast reactor corium melts containing steel or absorbers. Undesirable interactions were observed between the coating and these materials, leading to the carburization of the corium ingots. Concerning the resistance of the coating to oxide melts without ZrO 2, the zirconium carbide coating keeps its role of protective barrier with UO 2-Al 2O 3 below 2000 °C but does not resist to a UO 2-Eu 2O 3 mixture.

  14. Cooling System Design for PEM Fuel Cell Powered Air Vehicles

    Science.gov (United States)

    2010-06-18

    radiator #7. The fan blades and shroud were formed using stereo lithography; the fan motor was a brushless DC motor with motor controller. These...Naval Research Laboratory Washington, DC 20375-5320 NRL/MR/6110--10-9253 Cooling System Design for PEM Fuel Cell Powered Air Vehicles June 18, 2010...Stroman, Michael W. Schuette,* and Gregory S. Page† Naval Research Laboratory 4555 Overlook Avenue, SW Washington, DC 20375-5342 NRL/MR/6110--10-9253

  15. Engineering review of the core support structure of the Gas Cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-09-01

    The review of the core support structure of the gas cooled fast breeder reactor (GCFR) covered such areas as the design criteria, the design and analysis of the concepts, the development plan, and the projected manufacturing costs. Recommendations are provided to establish a basis for future work on the GCFR core support structure.

  16. Safe design of cooled tubular reactors for exothermic multiple reactions: Multiple-reaction networks

    NARCIS (Netherlands)

    Westerink, E.J.; Westerterp, K.R.

    1988-01-01

    The model of the pseudo-homogeneous, one-dimensional cooled tubular reactor is applied to a multiple-reaction network. It is demonstrated for a network which consists of two parallel and two consecutive reactions. Three criteria are developed to obtain an integral yield which does not deviate more t

  17. Gas-cooled fast reactor program. Progress report, January 1, 1980-June 30, 1981

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.

    1981-09-01

    Since the national Gas-Cooled Fast Breeder Reactor Program has been terminated, this document is the last progress report until reinstatement. It is divided into three sections: Core Flow Test Loop, GCFR shielding and physics, and GCFR pressure vessel and closure studies. (DLC)

  18. Thermally safe operation of a cooled semi-batch reactor: slow liquid-liquid reactions

    NARCIS (Netherlands)

    Steensma, M.; Westerterp, K.R.

    1988-01-01

    Thermally safe operation of a semi-batch reactor (SBR) implies that conditions leading to strong accumulation of unreacted reactants must be avoided. All thermal responses of a SBR, in which a slow liquid-liquid reaction takes place, can be represented in a diagram with the kinetics, cooling capacit

  19. Design guide for category II reactors light and heavy water cooled reactors. [US DOE

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-05-01

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

  20. A preliminary safety analysis for the prototype Gen IV Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kwi Lim; Ha, Kwi Seok; Jeong, Jae Ho; Choi, Chi Woong; Jeong, Tae Kyeong; Ahn, Sang June; Lee, Seung Won; Chang, Won Pyo; Kang, Seok Hun; Yoo, Jae Woon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Korea Atomic Energy Research Institute has been developing a pool-type sodium-cooled fast reactor of the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR). To assess the effectiveness of the inherent safety features of the PGSFR, the system transients during design basis accidents and design extended conditions are analyzed with MARS-LMR and the subchannel blockage events are analyzed with MATRA-LMR-FB. In addition, the in-vessel source term is calculated based on the super-safe, small, and simple reactor methodology. The results show that the PGSFR meets safety acceptance criteria with a sufficient margin during the events and keeps accidents from deteriorating into more severe accidents.

  1. 10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

    Science.gov (United States)

    2010-01-01

    ... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... Approvals § 50.46 Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium...

  2. Application of hafnium hydride control rod to large sodium cooled fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ikeda, Kazumi, E-mail: kazumi_ikeda@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Moriwaki, Hiroyuki, E-mail: hiroyuki_moriwaki@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Ohkubo, Yoshiyuki, E-mail: yoshiyuki_okubo@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 34-17, Jingumae 2-Chome, Shibuya-ku, Tokyo 150-0001 (Japan); Iwasaki, Tomohiko, E-mail: tomohiko.iwasaki@qse.tohoku.ac.jp [Department of Quantum Science and Energy Engineering, Tohoku University, Aoba, Aramaki, Aoba-ku, Sendai-shi, Miyagi-ken 980-8579 (Japan); Konashi, Kenji, E-mail: konashi@imr.tohoku.ac.jp [Institute for Materials Research, Tohoku University, Narita-cho, Oarai-machi, Higashi-Ibaraki-gun, Ibaraki-ken 311-1313 (Japan)

    2014-10-15

    Highlights: • Application of hafnium hydride control rod to large sodium cooled fast breeder reactor. • This paper treats application of an innovative hafnium hydride control rod to a large sodium cooled fast breeder reactor. • Hydrogen absorption triples the reactivity worth by neutron spectrum shift at H/Hf ratio of 1.3. • Lifetime of the control rod quadruples because produced daughters of hafnium isotopes are absorbers. • Nuclear and thermal hydraulic characteristics of the reactor are as good as or better than B-10 enriched boron carbide. - Abstract: This study treats the feasibility of long-lived hafnium hydride control rod in a large sodium-cooled fast breeder reactor by nuclear and thermal analyses. According to the nuclear calculations, it is found that hydrogen absorption of hafnium triples the reactivity by the neutron spectrum shift at the H/Hf ratio of 1.3, and a hafnium transmutation mechanism that produced daughters are absorbers quadruples the lifetime due to a low incineration rate of absorbing nuclides under irradiation. That is to say, the control rod can function well for a long time because an irradiation of 2400 EFPD reduces the reactivity by only 4%. The calculation also reveals that the hafnium hydride control rod can apply to the reactor in that nuclear and thermal characteristics become as good as or better than 80% B-10 enriched boron carbide. For example, the maximum linear heat rate becomes 3% lower. Owing to the better power distribution, the required flow rate decreases approximately by 1%. Consequently, it is concluded on desk analyses that the long lived hafnium hydride control rod is feasible in the large sodium-cooled fast breeder reactor.

  3. Pebble Bed Reactors Design Optimization Methods and their Application to the Pebble Bed Fluoride Salt Cooled High Temperature Reactor (PB-FHR)

    Science.gov (United States)

    Cisneros, Anselmo Tomas, Jr.

    The Fluoride salt cooled High temperature Reactor (FHR) is a class of advanced nuclear reactors that combine the robust coated particle fuel form from high temperature gas cooled reactors, direct reactor auxillary cooling system (DRACS) passive decay removal of liquid metal fast reactors, and the transparent, high volumetric heat capacitance liquid fluoride salt working fluids---flibe (33%7Li2F-67%BeF)---from molten salt reactors. This combination of fuel and coolant enables FHRs to operate in a high-temperature low-pressure design space that has beneficial safety and economic implications. In 2012, UC Berkeley was charged with developing a pre-conceptual design of a commercial prototype FHR---the Pebble Bed- Fluoride Salt Cooled High Temperature Reactor (PB-FHR)---as part of the Nuclear Energy University Programs' (NEUP) integrated research project. The Mark 1 design of the PB-FHR (Mk1 PB-FHR) is 236 MWt flibe cooled pebble bed nuclear heat source that drives an open-air Brayton combine-cycle power conversion system. The PB-FHR's pebble bed consists of a 19.8% enriched uranium fuel core surrounded by an inert graphite pebble reflector that shields the outer solid graphite reflector, core barrel and reactor vessel. The fuel reaches an average burnup of 178000 MWt-d/MT. The Mk1 PB-FHR exhibits strong negative temperature reactivity feedback from the fuel, graphite moderator and the flibe coolant but a small positive temperature reactivity feedback of the inner reflector and from the outer graphite pebble reflector. A novel neutronics and depletion methodology---the multiple burnup state methodology was developed for an accurate and efficient search for the equilibrium composition of an arbitrary continuously refueled pebble bed reactor core. The Burnup Equilibrium Analysis Utility (BEAU) computer program was developed to implement this methodology. BEAU was successfully benchmarked against published results generated with existing equilibrium depletion codes VSOP

  4. Transient air cooling thermal modeling of a PEM fuel cell

    Energy Technology Data Exchange (ETDEWEB)

    Adzakpa, K.P.; Ramousse, J.; Dube, Y.; Akremi, H.; Agbossou, K. [Hydrogen Research Institute and Departement de genie electrique, Universite du Quebec a Trois-Rivieres, CP 500, Trois-Rivieres (QC) (Canada); Dostie, M.; Poulin, A.; Fournier, M. [LTE-Hydro-Quebec, 600 av. de la Montagne, Shawinigan (QC) (Canada)

    2008-04-15

    Fuel cell utilization for automobile and residential applications is a promising option in order to help reduce environmental concerns such as pollution. However, fuel cell development requires addressing their dynamic behavior to improve their performances and their life cycle. Since the temperature distribution in the cell is known to be an important factor to the fuel cell's efficiency, a cooling device is often added to homogenize the temperature in the cell and to ensure temperature control. A 3D dynamic thermal model of a single fuel cell is presented in this work in order to study the temperature distribution in a fuel cell cooled from the bottom to the top with air. The model is governed by the thermal energy balance, taking into account the inlet gas humidity. The model is developed with the finite difference method and is implemented in the Matlab/Simulink environment. The validation is based on the performances of the ''NEXA'' fuel cell produced by Ballard Power Systems. The efficiency analysis of that air cooling device reveals that the cell temperature is directly linked to the current density and to the gas humidity - varying from 30 C at 5A to 80 C at 35A at low humidity. Moreover, the temperature non-uniformity in the stack is shown to be very high. As a result, temperatures are higher at the top part of the cell than at the bottom part, with a difference of up to a 5 C. Moreover the non-uniformity of the air cooling between the cells of the stack leads to large temperature variations, up to 8 C, from one cell to another. These temperature variations result in large voltage disparities between the cells, which reduce the total electrical power of the entire stack. (author)

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

    Science.gov (United States)

    Palmer, David J.

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

  6. Development of a plant dynamics computer code for analysis of a supercritical carbon dioxide Brayton cycle energy converter coupled to a natural circulation lead-cooled fast reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Moisseytsev, A.; Sienicki, J. J.

    2007-03-08

    STAR-LM is a lead-cooled pool-type fast reactor concept operating under natural circulation of the coolant. The reactor core power is 400 MWt. The open-lattice core consists of fuel pins attached to the core support plate, (the does not consist of removable fuel assemblies). The coolant flows outside of the fuel pins. The fuel is transuranic nitride, fabricated from reprocessed LWR spent fuel. The cladding material is HT-9 stainless steel; the steady-state peak cladding temperature is 650 C. The coolant is single-phase liquid lead under atmospheric pressure; the core inlet and outlet temperatures are 438 C and 578 C, respectively. (The Pb coolant freezing and boiling temperatures are 327 C and 1749 C, respectively). The coolant is contained inside of a reactor vessel. The vessel material is Type 316 stainless steel. The reactor is autonomous meaning that the reactor power is self-regulated based on inherent reactivity feedbacks and no external power control (through control rods) is utilized. The shutdown (scram) control rods are used for startup and shutdown and to stop the fission reaction in case of an emergency. The heat from the reactor is transferred to the S-CO{sub 2} Brayton cycle in in-reactor heat exchangers (IRHX) located inside the reactor vessel. The IRHXs are shell-and-tube type heat exchangers with lead flowing downwards on the shell side and CO{sub 2} flowing upwards on the tube side. No intermediate circuit is utilized. The guard vessel surrounds the reactor vessel to contain the coolant, in the very unlikely event of reactor vessel failure. The Reactor Vessel Auxiliary Cooling System (RVACS) implementing the natural circulation of air flowing upwards over the guard vessel is used to cool the reactor, in the case of loss of normal heat removal through the IRHXs. The RVACS is always in operation. The gap between the vessels is filled with liquid lead-bismuth eutectic (LBE) to enhance the heat removal by air by significantly reducing the thermal

  7. Laboratory study on the cooling effect of flash water evaporative cooling technology for ventilation and air-conditioning of buildings

    DEFF Research Database (Denmark)

    Fang, Lei; Yuan, Shu; Yang, Jianrong

    2016-01-01

    This paper presents a simple cooling technology using flash water evaporation. The technology combines a water atomizer with a plate heat exchanger used for heat recovery of a ventilation system. It is mainly used to cool the ventilation airflow from outdoors and is particularly suitable to be used...... in warm/hot and dry environment where dehumidification of outdoor air is not needed. A laboratory experiment was designed and conducted to evaluate the cooling effectiveness of this technology. The experiment was conducted in a twin-climate chamber. One chamber simulated warm/hot and dry outdoor...... environments and the other simulated an air-conditioned indoor environment. The flash water evaporation cooling device was installed in the chamber that simulated indoor environment. The air from the chamber simulating outdoor environment was introduced into the cooling device and cooled by the flash water...

  8. Experimental Study of the Effect of Graphite Dispersion on the Heat Transfer Phenomena in a Reactor Cavity Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Vaghetto, Rodolfo; Capone, Luigi; Hassan, Yassin A

    2011-05-31

    An experimental activity was performed to observe and study the effects of graphite dispersion and deposition on thermal-hydraulic phenomena in a reactor cavity cooling system (RCCS). The small-scale RCCS experimental facility (16.5 x 16.5 x 30.4 cm) used for this activity represents half of the reactor cavity with an electrically heated vessel. Water flowing through five vertical pipes removes the heat produced in the vessel and releases it into the environment by mixing with cold water in a large tank. The particle image velocimetry technique was used to study the velocity field of the air inside the cavity. A set of 52 thermocouples was installed in the facility to monitor the temperature profiles of the vessel, pipe walls, and air. Ten grams of a fine graphite powder (average particle size 2 m) was injected into the cavity through a spraying nozzle placed at the bottom of the vessel. The temperatures and air velocity field were recorded and compared with the measurements obtained before the graphite dispersion, showing a decrease of the temperature surfaces that was related to an increase in their emissivity. The results contribute to the understanding of RCCS capability in an accident scenario.

  9. A 50-100 kWe gas-cooled reactor for use on Mars.

    Energy Technology Data Exchange (ETDEWEB)

    Peters, Curtis D. (.)

    2006-04-01

    In the space exploration field there is a general consensus that nuclear reactor powered systems will be extremely desirable for future missions to the outer solar system. Solar systems suffer from the decreasing intensity of solar radiation and relatively low power density. Radioisotope Thermoelectric Generators are limited to generating a few kilowatts electric (kWe). Chemical systems are short-lived due to prodigious fuel use. A well designed 50-100 kWe nuclear reactor power system would provide sufficient power for a variety of long term missions. This thesis will present basic work done on a 50-100 kWe reactor power system that has a reasonable lifespan and would function in an extraterrestrial environment. The system will use a Gas-Cooled Reactor that is directly coupled to a Closed Brayton Cycle (GCR-CBC) power system. Also included will be some variations on the primary design and their effects on the characteristics of the primary design. This thesis also presents a variety of neutronics related calculations, an examination of the reactor's thermal characteristics, feasibility for use in an extraterrestrial environment, and the reactor's safety characteristics in several accident scenarios. While there has been past work for space reactors, the challenges introduced by thin atmospheres like those on Mars have rarely been considered.

  10. Modular High Temperature Gas-Cooled Reactor Safety Basis and Approach

    Energy Technology Data Exchange (ETDEWEB)

    David Petti; Jim Kinsey; Dave Alberstein

    2014-01-01

    Various international efforts are underway to assess the safety of advanced nuclear reactor designs. For example, the International Atomic Energy Agency has recently held its first Consultancy Meeting on a new cooperative research program on high temperature gas-cooled reactor (HTGR) safety. Furthermore, the Generation IV International Forum Reactor Safety Working Group has recently developed a methodology, called the Integrated Safety Assessment Methodology, for use in Generation IV advanced reactor technology development, design, and design review. A risk and safety assessment white paper is under development with respect to the Very High Temperature Reactor to pilot the Integrated Safety Assessment Methodology and to demonstrate its validity and feasibility. To support such efforts, this information paper on the modular HTGR safety basis and approach has been prepared. The paper provides a summary level introduction to HTGR history, public safety objectives, inherent and passive safety features, radionuclide release barriers, functional safety approach, and risk-informed safety approach. The information in this paper is intended to further the understanding of the modular HTGR safety approach. The paper gives those involved in the assessment of advanced reactor designs an opportunity to assess an advanced design that has already received extensive review by regulatory authorities and to judge the utility of recently proposed new methods for advanced reactor safety assessment such as the Integrated Safety Assessment Methodology.

  11. Comparison of In-Vessel Shielding Design Concepts between Sodium-cooled Fast Burner Reactor and the Sodium-cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    In this study, quantities of in-vessel shields were derived and compared each other based on the replaceable shield assembly concept for both of the breeder and burner SFRs. Korean Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) like SFR was used as the reference reactor and calculation method reported in the reference was used for shielding analysis. In this paper, characteristics of in-vessel shielding design were studied for the burner SFR and breeder SFR based on the replaceable shield assembly concept. An in-vessel shield to prevent secondary sodium activation (SSA) in the intermediate heat exchangers (IHXs) is one of the most important structures for the pool type Sodium-cooled Fast Reactor (SFR). In our previous work, two in-vessel shielding design concepts were compared each other for the burner SFR. However, a number of SFRs have been designed and operated with the breeder concept, in which axial and radial blankets were loaded for fuel breeding, during the past several decades. Since axial and radial blanket plays a role of neutron shield, comparison of required in-vessel shield amount between the breeder and burner SFRs may be an interesting work for SFR designer. Due to the blanket, the breeder SFR showed better performance in axial neutron shielding. Hence, 10.1 m diameter reactor vessel satisfied the design limit of SSA at the IHXs. In case of the burner SFR, due to more significant axial fast neutron leakage, 10.6 m diameter reactor vessel was required to satisfy the design limit of SSA at the IHXs. Although more efficient axial shied such as a mixture of ZrH{sub 2} and B{sub 4}C can improve shielding performance of the burner SFR, additional fabrication difficulty may mitigate the advantage of improved shielding performance. Therefore, it can be concluded that the breeder SFR has better characteristic in invessel shielding design to prevent SSA at the IHXs than the burner SFR in the pool-type reactor.

  12. Optimization of Internal Cooling Fins for Metal Hydride Reactors

    Directory of Open Access Journals (Sweden)

    Vamsi Krishna Kukkapalli

    2016-06-01

    Full Text Available Metal hydride alloys are considered as a promising alternative to conventional hydrogen storage cylinders and mechanical hydrogen compressors. Compared to storing in a classic gas tank, metal hydride alloys can store hydrogen at nearly room pressure and use less volume to store the same amount of hydrogen. However, this hydrogen storage method necessitates an effective way to reject the heat released from the exothermic hydriding reaction. In this paper, a finned conductive insert is adopted to improve the heat transfer in the cylindrical reactor. The fins collect the heat that is volumetrically generated in LaNi5 metal hydride alloys and deliver it to the channel located in the center, through which a refrigerant flows. A multiple-physics modeling is performed to analyze the transient heat and mass transfer during the hydrogen absorption process. Fin design is made to identify the optimum shape of the finned insert for the best heat rejection. For the shape optimization, use of a predefined transient heat generation function is proposed. Simulations show that there exists an optimal length for the fin geometry.

  13. Coupling a Supercritical Carbon Dioxide Brayton Cycle to a Helium-Cooled Reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Middleton, Bobby [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Pasch, James Jay [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Kruizenga, Alan Michael [Sandia National Lab. (SNL-CA), Livermore, CA (United States); Walker, Matthew [Sandia National Lab. (SNL-CA), Livermore, CA (United States)

    2016-01-01

    This report outlines the thermodynamics of a supercritical carbon dioxide (sCO2) recompression closed Brayton cycle (RCBC) coupled to a Helium-cooled nuclear reactor. The baseline reactor design for the study is the AREVA High Temperature Gas-Cooled Reactor (HTGR). Using the AREVA HTGR nominal operating parameters, an initial thermodynamic study was performed using Sandia's deterministic RCBC analysis program. Utilizing the output of the RCBC thermodynamic analysis, preliminary values of reactor power and of Helium flow rate through the reactor were calculated in Sandia's HelCO2 code. Some research regarding materials requirements was then conducted to determine aspects of corrosion related to both Helium and to sCO2 , as well as some mechanical considerations for pressures and temperatures that will be seen by the piping and other components. This analysis resulted in a list of materials-related research items that need to be conducted in the future. A short assessment of dry heat rejection advantages of sCO2> Brayton cycles was also included. This assessment lists some items that should be investigated in the future to better understand how sCO2 Brayton cycles and nuclear can maximally contribute to optimizing the water efficiency of carbon free power generation

  14. Preliminary Reactor Head Bolt Design of Prototype Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Han, Insu; Koo, Gyeonghoi [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    As structural requirements, the reactor head is designed to withstand all of the pressure, temperatures and forces which are likely to be imposed on it. The bolts that fasten the head to the vessel flange. Design of the reactor head bolts so as to withstand the loads applied should be designed. Currently, preliminary design of the PGSFR reactor bolts is progressed. So far, we have designed and evaluated example. The number and cross-sectional areas of bolts were determined using the procedure given in ASME BPVC Section III, Division 1, Appendix E. The purpose of this study is to conduct design the number and cross-sectional area of bolts attaching the PGSFR reactor head to the reactor vessel, using the ASME procedure. In this paper, preliminary bolt design for PGSFR was carried out according to the ASME procedure. Detailed calculations were carried out for bolt root diameter = 80 mm and number of bolts Nb = 45. It should be noted that the seating pressure recommended in the ASME code is only a suggested value, not mandatory appendix E. It does not guarantee a leak-tight joint. So these quantities are needed to carry out fatigue analysis of the bolts and to assure leak tightness of the joint during operation. For the future work, the fatigue and seismic analysis will be performed.

  15. A fundamentally new approach to air-cooled heat exchangers.

    Energy Technology Data Exchange (ETDEWEB)

    Koplow, Jeffrey P.

    2010-01-01

    We describe breakthrough results obtained in a feasibility study of a fundamentally new architecture for air-cooled heat exchangers. A longstanding but largely unrealized opportunity in energy efficiency concerns the performance of air-cooled heat exchangers used in air conditioners, heat pumps, and refrigeration equipment. In the case of residential air conditioners, for example, the typical performance of the air cooled heat exchangers used for condensers and evaporators is at best marginal from the standpoint the of achieving maximum the possible coefficient of performance (COP). If by some means it were possible to reduce the thermal resistance of these heat exchangers to a negligible level, a typical energy savings of order 30% could be immediately realized. It has long been known that a several-fold increase in heat exchanger size, in conjunction with the use of much higher volumetric flow rates, provides a straight-forward path to this goal but is not practical from the standpoint of real world applications. The tension in the market place between the need for energy efficiency and logistical considerations such as equipment size, cost and operating noise has resulted in a compromise that is far from ideal. This is the reason that a typical residential air conditioner exhibits significant sensitivity to reductions in fan speed and/or fouling of the heat exchanger surface. The prevailing wisdom is that little can be done to improve this situation; the 'fan-plus-finned-heat-sink' heat exchanger architecture used throughout the energy sector represents an extremely mature technology for which there is little opportunity for further optimization. But the fact remains that conventional fan-plus-finned-heat-sink technology simply doesn't work that well. Their primary physical limitation to performance (i.e. low thermal resistance) is the boundary layer of motionless air that adheres to and envelops all surfaces of the heat exchanger. Within this

  16. Validation of Reactor Physics-Thermal hydraulics Calculations for Research Reactors Cooled by the Laminar Flow of Water

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, K. A.; Schubring, D. [Univ. of Florida, Florida (United States); Girardin, G.; Pautz, A. [Swiss Federal Institute of Technology, Zuerich (Switzerland)

    2013-07-01

    A collaboration between the University of Florida and the Swiss Federal Institute of Technology, Lausanne (EPFL) has been formed to develop and validate detailed coupled multiphysics models of the zero-power (100 W) CROCUS reactor at EPFL and the 100 kW University of Florida Training Reactor, for the comprehensive analysis of the reactor behavior under transient (neutronic or thermal-hydraulic induced) conditions. These two reactors differ significantly in the core design and thermal power output, but share unique heat transfer and flow characteristics. They are characterized by single-phase laminar water flow at near-atmospheric pressures in complex geometries with the possibility of mechanically entrained air bubbles. Validation experiments will be designed to expand the validation domain of these existing models, computational codes and techniques. In this process, emphasis will be placed on validation of the coupled models developed to gain confidence in their applicability for safety analysis. EPFL is responsible for the design and implementation of transient experiments to generate a database of reactor parameters (flow distribution, power profile, and power evolution) to be used to validate against code predictions. The transient experiments performed at EPFL will be simulated on the basis of developed models for these tasks. Comparative analysis will be performed with SERPENT and MCNPX reference core models. UF focuses on the generation of the coupled neutron kinetics and thermal-hydraulic models, including implementation of a TRACE/PARCS reactor simulator model, a PARET model, and development of full-field computational fluid dynamics models (using OpenFOAM) for refined thermal-hydraulics physics treatments. In this subtask of the project, the aim is to verify by means of CFD the validity of TRACE predictions for near-atmospheric pressure water flow in the presence of mechanically entrained air bubbles. The scientific understanding of these multiphysics

  17. Evaluation of proposed German safety criteria for high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Barsell, A.W.

    1980-05-01

    This work reviews proposed safety criteria prepared by the German Bundesministerium des Innern (BMI) for future licensing of gas-cooled high-temperature reactor (HTR) concepts in the Federal Republic of Germany. Comparison is made with US General Design Criteria (GDCs) in 10CFR50 Appendix A and with German light water reactor (LWR) criteria. Implications for the HTR design relative to the US design and safety approach are indicated. Both inherent characteristics and design features of the steam cycle, gas turbine, and process heat concepts are taken into account as well as generic design options such as a pebble bed or prismatic core.

  18. Study on measure of cooling air in driving road with two air pipes

    Energy Technology Data Exchange (ETDEWEB)

    Wu, Q.; Qin, Y.; Zhai, M.; Zhang, J. [Heilongjiang Institute of Science and Technology, Jixi (China). Dept. of Resource and Environment Engineering

    2002-10-01

    If artificial refrigeration cooling is not available at a heading, then using the double-pipes air supply is an effective measure for reducing the air temperature at the face. Based on the characteristics of mutual control between the air inside and outside the pipes, the air temperature variation in the driving road with double-pipes installed is analysed. The method to calculate the air temperature is presented and the computer program based on this is designed. This program, as a reliable tool, can calculate the air temperature inside and outside the pipes. Based on the result of calculation for a heading face in Sanhejian Coal Mine of Xuzhou Mining Administration, the law of variaton of air temperature is analysed. 3 refs., 4 figs., 2 tabs.

  19. Assessing and controlling corrosion in air-cooled condensers

    Energy Technology Data Exchange (ETDEWEB)

    Dooley, R. Barry [Structural Integrity Associates, Inc., Charlotte, NC (United States); Preez, Francois du [Eskom (South Africa); Aspden, J. Denis; Howell, Andrew G.

    2009-05-15

    An increasing number of air-cooled condensers (ACC) are being installed and operated on conventional and combined cycle plants worldwide. Unless understood and corrected, the corrosion associated with the ACC ducts and tube entries can become a major problem for operators of plant. Up to just a few years ago very little was known about the corrosion/ flow-accelerated corrosion (FAC) process. This paper starts to rectify the situation with a description of the corrosion/FAC process, a corrosion index and a relationship between the operating pH and the level of iron at the condensate pump discharge. (orig.)

  20. Core Design and Deployment Strategy of Heavy Water Cooled Sustainable Thorium Reactor

    Directory of Open Access Journals (Sweden)

    Naoyuki Takaki

    2012-08-01

    Full Text Available Our previous studies on water cooled thorium breeder reactor based on matured pressurized water reactor (PWR plant technology concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array and using heavy water as coolant is appropriate for achieving better breeding performance and higher burn-up simultaneously [1–6]. One optimum core that produces 3.5 GW thermal energy using Th-233U oxide fuel shows a breeding ratio of 1.07 and averaged burn-up of about 80 GWd/t with long cycle length of 1300 days. The moderator to fuel volume ratio is 0.6 and required enrichment of 233U for the fresh fuel is about 7%. The coolant reactivity coefficient is negative during all cycles despite it being a large scale breeder reactor. In order to introduce this sustainable thorium reactor, three-step deployment scenario, with intermediate transition phase between current light water reactor (LWR phase and future sustainer phase, is proposed. Both in transition phase and sustainer phase, almost the same core design can be applicable only by changing fissile materials mixed with thorium from plutonium to 233U with slight modification in the fuel assembly design. Assuming total capacity of 60 GWe in current LWR phase and reprocessing capacity of 800 ton/y with further extensions to 1600 ton/y, all LWRs will be replaced by heavy water cooled thorium reactors within about one century then thorium reactors will be kept operational owing to its potential to sustain fissile fuels while reprocessing all spent fuels until exhaustion of massive thorium resource.

  1. Water cooled breeder program summary report (LWBR (Light Water Breeder Reactor) development program)

    Energy Technology Data Exchange (ETDEWEB)

    1987-10-01

    The purpose of the Department of Energy Water Cooled Breeder Program was to demonstrate pratical breeding in a uranium-233/thorium fueled core while producing electrical energy in a commercial water reactor generating station. A demonstration Light Water Breeder Reactor (LWBR) was successfully operated for more than 29,000 effective full power hours in the Shippingport Atomic Power Station. The reactor operated with an availability factor of 76% and had a gross electrical output of 2,128,943,470 kilowatt hours. Following operation, the expended core was examined and no evidence of any fuel element defects was found. Nondestructive assay of 524 fuel rods determined that 1.39 percent more fissile fuel was present at the end of core life than at the beginning, proving that breeding had occurred. This demonstrates the existence of a vast source of electrical energy using plentiful domestic thorium potentially capable of supplying the entire national need for many centuries. To build on the successful design and operation of the Shippingport Breeder Core and to provide the technology to implement this concept, several reactor designs of large breeders and prebreeders were developed for commercial-sized plants of 900--1000 Mw(e) net. This report summarizes the Water Cooled Breeder Program from its inception in 1965 to its completion in 1987. Four hundred thirty-six technical reports are referenced which document the work conducted as part of this program. This work demonstrated that the Light Water Breeder Reactor is a viable alternative as a PWR replacement in the next generation of nuclear reactors. This transition would only require a minimum of change in design and fabrication of the reactor and operation of the plant.

  2. Decay Heat Removal in GEN IV Gas-Cooled Fast Reactors

    Directory of Open Access Journals (Sweden)

    Lap-Yan Cheng

    2009-01-01

    Full Text Available The safety goal of the current designs of advanced high-temperature thermal gas-cooled reactors (HTRs is that no core meltdown would occur in a depressurization event with a combination of concurrent safety system failures. This study focused on the analysis of passive decay heat removal (DHR in a GEN IV direct-cycle gas-cooled fast reactor (GFR which is based on the technology developments of the HTRs. Given the different criteria and design characteristics of the GFR, an approach different from that taken for the HTRs for passive DHR would have to be explored. Different design options based on maintaining core flow were evaluated by performing transient analysis of a depressurization accident using the system code RELAP5-3D. The study also reviewed the conceptual design of autonomous systems for shutdown decay heat removal and recommends that future work in this area should be focused on the potential for Brayton cycle DHRs.

  3. Phytoplankton distribution in three thermally different but edaphically similar reactor cooling reservoirs

    Energy Technology Data Exchange (ETDEWEB)

    Wilde, E W

    1982-01-01

    Phytoplankton community structure and the physicochemical characteristics of three reactor cooling reservoirs in close proximity and of similar age and bottom type were studied during 1978. The three reservoirs differed in thermal alteration resulting from reactor cooling water as follows: (1) considerable heating with lake-wide temperatures >30/sup 0/C, even in winter; (2) a maximal 5/sup 0/C increase occurring in only one of three major arms of the reservoir; and (3) no thermal effluent received during the study period. Considerable spatial and temporal differences in water quality and phytoplankton community structure were observed; however, water temperature independent of other environmental factors (e.g., light and nutrients) was found to be a relatively unimportant variable for explaining phytoplankton periodicity.

  4. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for AHTRs

    Energy Technology Data Exchange (ETDEWEB)

    Lv, Quiping [The Ohio State Univ., Columbus, OH (United States); Sun, Xiaodong [The Ohio State Univ., Columbus, OH (United States); Chtistensen, Richard [The Ohio State Univ., Columbus, OH (United States); Blue, Thomas [The Ohio State Univ., Columbus, OH (United States); Yoder, Graydon [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-05-08

    The principal objective of this research is to test and model the heat transfer performance and reliability of the Direct Reactor Auxiliary Cooling System (DRACS) for AHTRs. In addition, component testing of fluidic diodes is to be performed to examine the performance and viability of several existing fluidic diode designs. An extensive database related to the thermal performance of the heat exchangers involved will be obtained, which will be used to benchmark a computer code for the DRACS design and to evaluate and improve, if needed, existing heat transfer models of interest. The database will also be valuable for assessing the viability of the DRACS concept and benchmarking any related computer codes in the future. The experience of making a liquid fluoride salt test facility available, with lessons learned, will greatly benefit the development of the Fluoride Salt-cooled High-temperature Reactor (FHR) and eventually the AHTR programs.

  5. An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Cetiner, Sacit M [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

    2009-11-01

    This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.

  6. Monte Carlo modeling of Lead-Cooled Fast Reactor in adiabatic equilibrium state

    Energy Technology Data Exchange (ETDEWEB)

    Stanisz, Przemysław, E-mail: pstanisz@agh.edu.pl; Oettingen, Mikołaj, E-mail: moettin@agh.edu.pl; Cetnar, Jerzy, E-mail: cetnar@mail.ftj.agh.edu.pl

    2016-05-15

    Graphical abstract: - Highlights: • We present the Monte Carlo modeling of the LFR in the adiabatic equilibrium state. • We assess the adiabatic equilibrium fuel composition using the MCB code. • We define the self-adjusting process of breeding gain by the control rod operation. • The designed LFR can work in the adiabatic cycle with zero fuel breeding. - Abstract: Nuclear power would appear to be the only energy source able to satisfy the global energy demand while also achieving a significant reduction of greenhouse gas emissions. Moreover, it can provide a stable and secure source of electricity, and plays an important role in many European countries. However, nuclear power generation from its birth has been doomed by the legacy of radioactive nuclear waste. In addition, the looming decrease in the available resources of fissile U235 may influence the future sustainability of nuclear energy. The integrated solution to both problems is not trivial, and postulates the introduction of a closed-fuel cycle strategy based on breeder reactors. The perfect choice of a novel reactor system fulfilling both requirements is the Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state. In such a state, the reactor converts depleted or natural uranium into plutonium while consuming any self-generated minor actinides and transferring only fission products as waste. We present the preliminary design of a Lead-Cooled Fast Reactor operating in the adiabatic equilibrium state with the Monte Carlo Continuous Energy Burnup Code – MCB. As a reference reactor model we apply the core design developed initially under the framework of the European Lead-cooled SYstem (ELSY) project and refined in the follow-up Lead-cooled European Advanced DEmonstration Reactor (LEADER) project. The major objective of the study is to show to what extent the constraints of the adiabatic cycle are maintained and to indicate the phase space for further improvements. The analysis

  7. A modular gas-cooled cermet reactor system for planetary base power

    Science.gov (United States)

    Jahshan, Salim N.; Borkowski, Jeffrey A.

    1993-01-01

    Fission nuclear power is foreseen as the source for electricity in planetary colonization and exploration. A six module gas-cooled, cermet-fueled reactor is proposed that can meet the design objectives. The highly enriched core is compact and can operate at high temperature for a long life. The helium coolant powers six modular Brayton cycles that compare favorably with the SP-100-based Brayton cycle.

  8. Thermally safe operation of a cooled semi-batch reactor: slow liquid-liquid reactions

    OpenAIRE

    Steensma, M.; Westerterp, K R

    1988-01-01

    Thermally safe operation of a semi-batch reactor (SBR) implies that conditions leading to strong accumulation of unreacted reactants must be avoided. All thermal responses of a SBR, in which a slow liquid-liquid reaction takes place, can be represented in a diagram with the kinetics, cooling capacity and potential temperature rise as the keyfactors. Slow reactions taking place in the dispersed phase were found to be more prone to accumulation than reactions in the continuous phase. An overhea...

  9. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  10. Simulation and control of water-gas shift packed bed reactor with inter-stage cooling

    Science.gov (United States)

    Saw, S. Z.; Nandong, J.

    2016-03-01

    Water-Gas Shift Reaction (WGSR) has become one of the well-known pathways for H2 production in industries. The issue with WGSR is that it is kinetically favored at high temperatures but thermodynamically favored at low temperatures, thus requiring careful consideration in the control design in order to ensure that the temperature used does not deactivate the catalyst. This paper studies the effect of a reactor arrangement with an inter-stage cooling implemented in the packed bed reactor to look at its effect on outlet temperature. A mathematical model is developed based on one-dimensional heat and mass transfers which incorporate the intra-particle effects. It is shown that the placement of the inter-stage cooling and the outlet temperature exiting the inter-stage cooling have strong influence on the reaction conversion. Several control strategies are explored for the process. It is shown that a feedback- feedforward control strategy using Multi-scale Control (MSC) is effective to regulate the reactor temperature profile which is critical to maintaining the catalysts activity.

  11. Impingement heat sinks for air cooled high power electronic modules

    Energy Technology Data Exchange (ETDEWEB)

    Kang, S.S.; Holahan, M.F. [IBM Corp., Rochester, MN (United States)

    1995-12-31

    The subject of the present work is a parallel plate heat sink that is designed so that the air flow impinges at the fin tips and exhausts over the two open side faces. This type of design attempts to achieve an air flow direction that is substantially opposite to the heat flow direction within the fins so as to exploit the greater heat transfer effectiveness of counterflow heat exchange. A one dimensional model of the heat sink was developed with the assumption of air flow from the fin tips to the fin base. This simplified model was used to identify an initial heat sink geometry to cool a specific multichip module. Computational Fluid Dynamics models that account for the actual flow pattern within the heat sink were used to study a range of variations to the initial geometry and to identify the best geometry over the range examined. Experimental heat transfer and pressure drop data is reported for two heat sink prototypes. The test data is in good agreement with CFD predictions. Suitable correlations for the heat sink thermal resistance and pressure drop versus the air flow rate are developed. The developed heat sink demonstrated an area specific thermal resistance better than 8.7 C (W/cm{sup 2}).

  12. Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

    Energy Technology Data Exchange (ETDEWEB)

    Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

    1982-01-01

    During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor.

  13. Pebble Fuel Handling and Reactivity Control for Salt-Cooled High Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Peterson, Per [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering; Greenspan, Ehud [Univ. of California, Berkeley, CA (United States). Dept. of Nuclear Engineering

    2015-02-09

    This report documents the work completed on the X-PREX facility under NEUP Project 11- 3172. This project seeks to demonstrate the viability of pebble fuel handling and reactivity control for fluoride salt-cooled high-temperature reactors (FHRs). The research results also improve the understanding of pebble motion in helium-cooled reactors, as well as the general, fundamental understanding of low-velocity granular flows. Successful use of pebble fuels in with salt coolants would bring major benefits for high-temperature reactor technology. Pebble fuels enable on-line refueling and operation with low excess reactivity, and thus simpler reactivity control and improved fuel utilization. If fixed fuel designs are used, the power density of salt- cooled reactors is limited to 10 MW/m3 to obtain adequate duration between refueling, but pebble fuels allow power densities in the range of 20 to 30 MW/m3. This can be compared to the typical modular helium reactor power density of 5 MW/m3. Pebble fuels also permit radial zoning in annular cores and use of thorium or graphite pebble blankets to reduce neutron fluences to outer radial reflectors and increase total power production. Combined with high power conversion efficiency, compact low-pressure primary and containment systems, and unique safety characteristics including very large thermal margins (>500°C) to fuel damage during transients and accidents, salt-cooled pebble fuel cores offer the potential to meet the major goals of the Advanced Reactor Concepts Development program to provide electricity at lower cost than light water reactors with improved safety and system performance.This report presents the facility description, experimental results, and supporting simulation methods of the new X-Ray Pebble Recirculation Experiment (X-PREX), which is now operational and being used to collect data on the behavior of slow dense granular flows relevant to pebble bed reactor core designs. The X

  14. Fluoride Salt-Cooled High-Temperature Reactor Technology Development and Demonstration Roadmap

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Flanagan, George F [ORNL; Mays, Gary T [ORNL; Pointer, William David [ORNL; Robb, Kevin R [ORNL; Yoder Jr, Graydon L [ORNL

    2013-11-01

    Fluoride salt-cooled High-temperature Reactors (FHRs) are an emerging reactor class with potentially advantageous performance characteristics, and fully passive safety. This roadmap describes the principal remaining FHR technology challenges and the development path needed to address the challenges. This roadmap also provides an integrated overview of the current status of the broad set of technologies necessary to design, evaluate, license, construct, operate, and maintain FHRs. First-generation FHRs will not require any technology breakthroughs, but do require significant concept development, system integration, and technology maturation. FHRs are currently entering early phase engineering development. As such, this roadmap is not as technically detailed or specific as would be the case for a more mature reactor class. The higher cost of fuel and coolant, the lack of an approved licensing framework, the lack of qualified, salt-compatible structural materials, and the potential for tritium release into the environment are the most obvious issues that remain to be resolved.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ohashi, Hirofumi; Sato, Hiroyuki; Tachibana, Yukio; Kunitomi, Kazuhiko; Ogawa, Masuro [Nuclear Hydrogen and Heat Application Research Center, Japan Atomic Energy Agency, Oarai-machi, Ibaraki-ken, 311-1394 (Japan)

    2012-06-06

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

  16. Modeling and energy simulation of the variable refrigerant flow air conditioning system with water-cooled condenser under cooling conditions

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yueming; Wu, Jingyi [Shanghai Jiao Tong University, Institute of Refrigeration and Cryogenics (China); Shiochi, Sumio [Daikin Industries Ltd. (Japan)

    2009-09-15

    As a new system, variable refrigerant flow system with water-cooled condenser (water-cooled VRF) can offer several interesting characteristics for potential users. However, at present, its dynamic simulation simultaneously in association with building and other equipments is not yet included in the energy simulation programs. Based on the EnergyPlus's codes, and using manufacturer's performance parameters and data, the special simulation module for water-cooled VRF is developed and embedded in the software of EnergyPlus. After modeling and testing the new module, on the basis of a typical office building in Shanghai with water-cooled VRF system, the monthly and seasonal cooling energy consumption and the breakdown of the total power consumption are analyzed. The simulation results show that, during the whole cooling period, the fan-coil plus fresh air (FPFA) system consumes about 20% more power than the water-cooled VRF system does. The power comparison between the water-cooled VRF system and the air-cooled VRF system is performed too. All of these can provide designers some ideas to analyze the energy features of this new system and then to determine a better scheme of the air conditioning system. (author)

  17. Development of Safety Analysis Codes and Experimental Validation for a Very High Temperature Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chang, H. Oh, PhD; Cliff Davis; Richard Moore

    2004-11-01

    The very high temperature gas-cooled reactors (VHTGRs) are those concepts that have average coolant temperatures above 900 degrees C or operational fuel temperatures above 1250 degrees C. These concepts provide the potential for increased energy conversion efficiency and for high-temperature process heat application in addition to power generation and nuclear hydrogen generation. While all the High Temperature Gas Cooled Reactor (HTGR) concepts have sufficiently high temperatures to support process heat applications, such as desalination and cogeneration, the VHTGR's higher temperatures are suitable for particular applications such as thermochemical hydrogen production. However, the high temperature operation can be detrimental to safety following a loss-of-coolant accident (LOCA) initiated by pipe breaks caused by seismic or other events. Following the loss of coolant through the break and coolant depressurization, air from the containment will enter the core by molecular diffusion and ultimately by natural convection, leading to oxidation of the in-core graphite structures and fuel. The oxidation will release heat and accelerate the heatup of the reactor core. Thus, without any effective countermeasures, a pipe break may lead to significant fuel damage and fission product release. The Idaho National Engineering and Environmental Laboratory (INEEL) has investigated this event for the past three years for the HTGR. However, the computer codes used, and in fact none of the world's computer codes, have been sufficiently developed and validated to reliably predict this event. New code development, improvement of the existing codes, and experimental validation are imperative to narrow the uncertaninty in the predictions of this type of accident. The objectives of this Korean/United States collaboration are to develop advanced computational methods for VHTGR safety analysis codes and to validate these computer codes.

  18. Emergency reactor core cooling water injection device for light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oda, Junro.

    1994-05-13

    A reactor pressure vessel is immersed in pool water of a reactor container. A control valve is interposed to a water supplying pipelines connecting pool water and a pressure vessel. A valve actuation means for opening/closing the control valve comprises a lifting tank. The inner side of the lifting tank and the inner side of the pressure vessel are connected by a communication pipeline (a syphon pipe) at upper and lower two portions. The lifting tank and the control valve are connected by a link mechanism. When a water level in the pressure vessel is lowered, the water level in the lifting tank is lowered to the same level as that in the pressure vessel. This reduces the weight of the lifting tank, the lifting tank is raised, to open the control valve by way of a link mechanism. As a result, liquid phase in the pressure vessel is in communication with the pool water, and the pool water flows down into the pressure vessel to maintain the reactor core in a flooded state. (I.N.).

  19. Development of level-1 PSA method applicable to Japan Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kurisaka, K., E-mail: kurisaka.kennichi@jaea.go.jp [Advanced Nuclear System R and D Directorate, Japan Atomic Energy Agency, Ibaraki (Japan); Sakai, T.; Yamano, H. [Advanced Nuclear System R and D Directorate, Japan Atomic Energy Agency, Ibaraki (Japan); Fujita, S.; Minagawa, K. [Department of Mechanical Engineering, School of Engineering, Tokyo Denki University, Tokyo (Japan); Yamaguchi, A.; Takata, T. [Department of Energy and Environment Engineering, Osaka University, Osaka (Japan)

    2014-04-01

    This paper describes a study to develop the level-1 probabilistic safety assessment (PSA) method that is applicable to the Japan Sodium-cooled Fast Reactor (JSFR). This study has been started since August 2010 and aims to provide a new evaluation method of (1) passive safety architectures related to internal events and (2) an advanced seismic isolation system related to a seismic event as a representative external event in Japan. Regarding the internal events evaluation, a quantitative analysis on the frequency of the core damage caused by reactor shutdown failure was conducted. A failure in passive reactor shutdown was taken into account in the event tree model. The failure rate of sodium-cooled fast reactor (SFR) specific components was evaluated based on the operating experience in existing SFRs by applying the Hierarchical Bayesian Method, which can consider a plant-to-plant variability. By conducting an uncertainty analysis, it was found that the assumption about the correlation of the probability parameters between the main and backup reactor shutdown systems (RSSs) is sensitive to the mean value of the frequency of the core damage caused by reactor shutdown failure. As for the seismic event evaluation, seismic response analysis and sensitivity analysis of a seismic isolation system were carried out. Rubber bearings have a hardening property in horizontal direction and a softening property in vertical direction in case of large deformation. Therefore the analyses considered nonlinearity of rubber bearings. Both horizontal and vertical nonlinear characteristics of rubber bearings were explained by multi-linear model. Mass point analytical models were applied. At first, seismic response analysis was executed in order to investigate influence of nonlinearity of rubber bearing upon response of building. Then sensitivity analysis was executed. Parameters of rubber bearings, oil dampers and the building were fluctuated, and influence of dispersion of these

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-07-15

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

  1. Reliability and safety of the K Reactor cooling system: Part 2, Engineering analysis of hydraulic and mechanical aspects

    Energy Technology Data Exchange (ETDEWEB)

    Shoemaker, R.H.

    1960-04-04

    Subsequent to the recent formulation and adoption of safety criteria for reactor cooling systems, there appeared the need for an independent evaluation of the safety and reliability of the K-Reactor cooling system in terms of these criteria. The primary, secondary and last-ditch cooling systems of this reactor involve a strong inter-dependence between electrical and hydraulic components of the water plant. Because of the complexity of inter-relationships between these components, the analysis was divided into two parallel studies which were accomplished during the simmer of 1959. F. D. Robbins has presented his analysis of the electrical power and control system in HW-61887. This report deals with an engineering analysis of the hydraulic and mechanical aspects of the reliability and safety of the K-Reactor Cooling System. The system, as described in this report, is that which existed during the simmer of 1959, prior to modification under Project CG-775 (now Project CG-883).

  2. Study of evaluation methods for in-vessel corium retention through external vessel cooling and safety of reactor cavity

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hoon; Chang, Soon Heung; Choi, Tae Hoon; Kim, Hyun Sop; Yang, Soo Hyung; Kim, Soo Hyung; Kim, Seung Hop; An Hyung Taek; Jeong, Yong Hoon; Huh, Gyun Young [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-03-15

    Cooling methodologies for the molten corium resulted from the severe accident of the nuclear power plant is suggested as one of most important items for the safety of the NPP. In this regard, considerable experimental and analytical works have been devoted. In the 1st phase of this project, present status related to the external reactor vessel cooling for the retention of the corium in the reactor vessel and corium at the reactor cavity have been investigated and preliminary studies have been accomplished for the detail evaluation of the each cooling methodology. The preliminary studies include the analysis and detail investigation of the possible phenomena, investigation of the heat transfer correlations and preliminary evaluation of the external reactor vessel cooling using the developed computer code.

  3. A review of existing gas-cooled reactor circulators with application of the lessons learned to the new production reactor circulators

    Energy Technology Data Exchange (ETDEWEB)

    White, L.S.

    1990-07-01

    This report presents the results of a study of the lessons learned during the design, testing, and operation of gas-cooled reactor coolant circulators. The intent of this study is to identify failure modes and problem areas of the existing circulators so this information can be incorporated into the design of the circulators for the New Production Reactor (NPR)-Modular High-Temperature Gas Cooled Reactor (MHTGR). The information for this study was obtained primarily from open literature and includes data on high-pressure, high-temperature helium test loop circulators as well as the existing gas cooled reactors worldwide. This investigation indicates that trouble free circulator performance can only be expected when the design program includes a comprehensive prototypical test program, with the results of this test program factored into the final circulator design. 43 refs., 7 tabs.

  4. Building on knowledge base of sodium cooled fast spectrum reactors to develop materials technology for fusion reactors

    Science.gov (United States)

    Raj, Baldev; Rao, K. Bhanu Sankara

    2009-04-01

    The alloys 316L(N) and Mod. 9Cr-1Mo steel are the major structural materials for fabrication of structural components in sodium cooled fast reactors (SFRs). Various factors influencing the mechanical behaviour of these alloys and different modes of deformation and failure in SFR systems, their analysis and the simulated tests performed on components for assessment of structural integrity and the applicability of RCC-MR code for the design and validation of components are highlighted. The procedures followed for optimal design of die and punch for the near net shape forming of petals of main vessel of 500 MWe prototype fast breeder reactor (PFBR); the safe temperature and strain rate domains established using dynamic materials model for forming of 316L(N) and 9Cr-1Mo steels components by various industrial processes are illustrated. Weldability problems associated with 316L(N) and Mo. 9Cr-1Mo are briefly discussed. The utilization of artificial neural network models for prediction of creep rupture life and delta-ferrite in austenitic stainless steel welds is described. The usage of non-destructive examination techniques in characterization of deformation, fracture and various microstructural features in SFR materials is briefly discussed. Most of the experience gained on SFR systems could be utilized in developing science and technology for fusion reactors. Summary of the current status of knowledge on various aspects of fission and fusion systems with emphasis on cross fertilization of research is presented.

  5. Tritium permeation behavior through pyrolytic carbon in tritium production using high-temperature gas-cooled reactor for fusion reactors

    Directory of Open Access Journals (Sweden)

    H. Ushida

    2016-12-01

    Full Text Available Under tritium production method using a high-temperature gas-cooled reactor loaded Li compound, Li compound has to be coated by ceramic materials in order to suppress the spreading of tritium to the whole reactor. Pyrolytic carbon (PyC is a candidate of the coating material because of its high resistance for gas permeation. In this study, hydrogen permeation experiments using a PyC-coated isotropic graphite tube were conducted and hydrogen diffusivity, solubility and permeability were evaluated. Tritium permeation behavior through PyC-coated Li compound particles was simulated by using obtained data. Hydrogen permeation flux through PyC in a steady state is proportional to the hydrogen pressure and is larger than that through Al2O3 which is also candidate coating material. However, total tritium leak within the supposed reactor operation period through the PyC-coated Li compound particles is lower than that through the Al2O3-coated ones because the hydrogen absorption capacity in PyC is considerably larger than that in Al2O3.

  6. On the performance of air conditioner with heat pipe for cooling air in the condenser

    Energy Technology Data Exchange (ETDEWEB)

    Naphon, Paisarn, E-mail: paisarnn@swu.ac.t [Thermo-Fluids and Heat Transfer Enhancement Laboratory (TFHT), Department of Mechanical Engineering, Faculty of Engineering, Srinakharinwirot University, 63 Rangsit-Nakhornnayok Rd., Ongkharak, Nakhorn-Nayok 26120 (Thailand)

    2010-11-15

    Improvement of the air conditioning system performance by using the heat pipe for cooling air before entering the condenser is presented. In the experiment, the heat pipe is fabricated from the straight copper tube with the diameter and length of 10, 600 mm, respectively. The arrangements of the heat pipe sets are arranged in the staggered layout with the tube rows of 1, 2, 3. R134a refrigerant is used as working fluid in the heat pipe set for this present study. By comparing with a conventional air conditioning system, the air conditioning system with three rows of heat pipe gives the highest COP and EER with increasing of 6.4%, 17.5%, respectively. On the global warming and environment problems, the results of this study are expected to lead to guidelines that will allow the improved performance of the air conditioning systems which reduce its energy consumption.

  7. Design study of lead bismuth cooled fast reactors and capability of natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Oktamuliani, Sri, E-mail: srioktamuliani@ymail.com; Su’ud, Zaki, E-mail: szaki@fi.itb.ac.id [Nuclear and Reactor Physics Laboratory, FMIPA, ITB, Physics Buildings, Jl. Ganesha 10, Bandung 40132 (Indonesia)

    2015-09-30

    A preliminary study designs SPINNOR (Small Power Reactor, Indonesia, No On-Site Refueling) liquid metal Pb-Bi cooled fast reactors, fuel (U, Pu)N, 150 MWth have been performed. Neutronic calculation uses SRAC which is designed cylindrical core 2D (R-Z) 90 × 135 cm, on the core fuel composed of heterogeneous with percentage difference of PuN 10, 12, 13% and the result of calculation is effective neutron multiplication 1.0488. Power density distribution of the output SRAC is generated for thermal hydraulic calculation using Delphi based on Pascal language that have been developed. The research designed a reactor that is capable of natural circulation at inlet temperature 300 °C with variation of total mass flow rate. Total mass flow rate affect pressure drop and temperature outlet of the reactor core. The greater the total mass flow rate, the smaller the outlet temperature, but increase the pressure drop so that the chimney needed more higher to achieve natural circulation or condition of the system does not require a pump. Optimization of the total mass flow rate produces optimal reactor design on the total mass flow rate of 5000 kg/s with outlet temperature 524,843 °C but require a chimney of 6,69 meters.

  8. Validation of SCALE for High Temperature Gas-Cooled Reactors Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [ORNL; Ilas, Dan [ORNL; Kelly, Ryan P [ORNL; Sunny, Eva E [ORNL

    2012-08-01

    This report documents verification and validation studies carried out to assess the performance of the SCALE code system methods and nuclear data for modeling and analysis of High Temperature Gas-Cooled Reactor (HTGR) configurations. Validation data were available from the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhE Handbook), prepared by the International Reactor Physics Experiment Evaluation Project, for two different HTGR designs: prismatic and pebble bed. SCALE models have been developed for HTTR, a prismatic fuel design reactor operated in Japan and HTR-10, a pebble bed reactor operated in China. The models were based on benchmark specifications included in the 2009, 2010, and 2011 releases of the IRPhE Handbook. SCALE models for the HTR-PROTEUS pebble bed configuration at the PROTEUS critical facility in Switzerland have also been developed, based on benchmark specifications included in a 2009 IRPhE draft benchmark. The development of the SCALE models has involved a series of investigations to identify particular issues associated with modeling the physics of HTGRs and to understand and quantify the effect of particular modeling assumptions on calculation-to-experiment comparisons.

  9. Conceptual System Design of a Supercritical CO2 cooled Micro Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seong Gu; Oh, Bongseong; Baik, Seung Joon; Yu, Hwanyeal; Kim, Yonghee; Lee, Jeong Ik [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    The S-CO2 Brayton cycle has many advantages for SMR's power conversion system. The S-CO2 cycle can achieve small component size and simple cycle layout as shown in Fig. 1. Therefore, a concept of one module containing the S-CO2 cooled fast reactor core and power conversion system is realizable. Thanks to the compact heat exchanger technology such as Printed Circuit Heat Exchanger (PCHE), the supercritical fluid with mediocre heat transfer performance can be utilized to a thermal cycle. This concept of fully modularized reactor is named as KAIST Micro Modular Reactor (MMR). It can achieve large economic by production in series, and transported in the land way or sea way. Based on the design results and dimensions of the reactor core and cycle components, the authors propose a conceptual layout of KAIST MMR. Based on this concept of reactor core, power conversion system, and decay heat removal system, the seasonal operation and transient analysis will be performed in the further works.

  10. Nonlinear Dynamic Modeling and Simulation of a Passively Cooled Small Modular Reactor

    Science.gov (United States)

    Arda, Samet Egemen

    A nonlinear dynamic model for a passively cooled small modular reactor (SMR) is developed. The nuclear steam supply system (NSSS) model includes representations for reactor core, steam generator, pressurizer, hot leg riser and downcomer. The reactor core is modeled with the combination of: (1) neutronics, using point kinetics equations for reactor power and a single combined neutron group, and (2) thermal-hydraulics, describing the heat transfer from fuel to coolant by an overall heat transfer resistance and single-phase natural circulation. For the helical-coil once-through steam generator, a single tube depiction with time-varying boundaries and three regions, i.e., subcooled, boiling, and superheated, is adopted. The pressurizer model is developed based upon the conservation of fluid mass, volume, and energy. Hot leg riser and downcomer are treated as first-order lags. The NSSS model is incorporated with a turbine model which permits observing the power with given steam flow, pressure, and enthalpy as input. The overall nonlinear system is implemented in the Simulink dynamic environment. Simulations for typical perturbations, e.g., control rod withdrawal and increase in steam demand, are run. A detailed analysis of the results show that the steady-state values for full power are in good agreement with design data and the model is capable of predicting the dynamics of the SMR. Finally, steady-state control programs for reactor power and pressurizer pressure are also implemented and their effect on the important system variables are discussed.

  11. Heat transfer and core neutronics considerations of the heat pipe cooled thermionic reactor

    Science.gov (United States)

    Determan, W. R.; Lewis, Brian

    1991-01-01

    The authors summarize the results of detailed neutronic and thermal-hydraulic evaluations of the heat pipe cooled thermionic (HPTI) reactor design, identify its key design attributes, and quantify its performance characteristics. The HPTI core uses modular, liquid-metal core heat transfer assemblies to replace the liquid-metal heat transport loop employed by in-core thermionic reactor designs of the past. The nuclear fuel, power conversion, heat transport, and heat rejection functions are all combined into a single modular unit. The reactor/converter assembly uses UN fuel pins to obtain a critical core configuration with in-core safety rods and reflector controls added to complete the subassembly. By thermally bonding the core heat transfer assemblies during the reactor core is coupled neutronically, thermally, and electrically into a modular assembly of individual power sources with cross-tied architecture. A forward-facing heat pipe radiator assembly extends from the reactor head in the shape of a frustum of a cone on the opposite side of the power system from the payload. Important virtues of the concept are the absence of any single-point failures and the ability of the core to effectively transfer the TFE waste heat load laterally to other in-core heat transfer assemblies in the event of multiple failures in either in-core and radiator heat pipes.

  12. Vibration test on KMRR reactor structure and primary cooling system piping

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1994-10-01

    Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author).

  13. A Compact Gas-Cooled Fast Reactor with an Ultra-Long Fuel Cycle

    Directory of Open Access Journals (Sweden)

    Hangbok Choi

    2013-01-01

    Full Text Available In an attempt to allow nuclear power to reach its full economic potential, General Atomics is developing the Energy Multiplier Module (EM2, which is a compact gas-cooled fast reactor (GFR. The EM2 augments its fissile fuel load with fertile materials to enhance an ultra-long fuel cycle based on a “convert-and-burn” core design which converts fertile material to fissile fuel and burns it in situ over a 30-year core life without fuel supplementation or shuffling. A series of reactor physics trade studies were conducted and a baseline core was developed under the specific physics design requirements of the long-life small reactor. The EM2 core performance was assessed for operation time, fuel burnup, excess reactivity, peak power density, uranium utilization, etc., and it was confirmed that an ultra-long fuel cycle core is feasible if the conversion is enough to produce fissile material and maintain criticality, the amount of matrix material is minimized not to soften the neutron spectrum, and the reactor core size is optimized to minimize the neutron loss. This study has shown the feasibility, from the reactor physics standpoint, of a compact GFR that can meet the objectives of ultra-long fuel cycle, factory-fabrication, and excellent fuel utilization.

  14. Heat transfer and core neutronics considerations of the heat pipe cooled thermionic reactor

    Science.gov (United States)

    Determan, W. R.; Lewis, Brian

    The authors summarize the results of detailed neutronic and thermal-hydraulic evaluations of the heat pipe cooled thermionic (HPTI) reactor design, identify its key design attributes, and quantify its performance characteristics. The HPTI core uses modular, liquid-metal core heat transfer assemblies to replace the liquid-metal heat transport loop employed by in-core thermionic reactor designs of the past. The nuclear fuel, power conversion, heat transport, and heat rejection functions are all combined into a single modular unit. The reactor/converter assembly uses UN fuel pins to obtain a critical core configuration with in-core safety rods and reflector controls added to complete the subassembly. By thermally bonding the core heat transfer assemblies during the reactor core is coupled neutronically, thermally, and electrically into a modular assembly of individual power sources with cross-tied architecture. A forward-facing heat pipe radiator assembly extends from the reactor head in the shape of a frustum of a cone on the opposite side of the power system from the payload. Important virtues of the concept are the absence of any single-point failures and the ability of the core to effectively transfer the TFE waste heat load laterally to other in-core heat transfer assemblies in the event of multiple failures in either in-core and radiator heat pipes.

  15. Design, Testing and Modeling of the Direct Reactor Auxiliary Cooling System for FHRs

    Science.gov (United States)

    Lu, Qiuping

    Direct Reactor Auxiliary Cooling System (DRACS) is a passive decay heat removal system proposed for the Fluoride-salt-cooled High-temperature Reactor (FHR) that combines coated particle fuel and a graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three coupled natural circulation/convection loops, relying completely on buoyancy as the driving force. These loops are coupled through two heat exchangers, namely, the DRACS Heat Exchanger (DHX) and the Natural Draft Heat Exchanger (NDHX). In addition, a fluidic diode is employed to minimize the parasitic flow into the DRACS primary loop and correspondingly the heat loss to the DRACS during normal operation of the reactor, but to keep the DRACS ready for activation, if needed, during accidents. While the DRACS concept has been proposed, there are no actual prototypic DRACS systems for FHRs built or tested in the literature. The primary goal of the present research is to design, test, and model the DRACS for FHR applications. Previously, a detailed modular design of the DRACS for a 20-MWth FHR was developed. As a starting point, the DRACS was designed to remove 1% of the reactor nominal power, i.e., 200 kW decay power. In addition, a detailed scaling analysis has been performed to develop the key non-dimensional numbers that characterize the DRACS system. Based on the previous work on the prototypic DRACS design and scaling analysis, two scaled-down test facilities have been designed and constructed, namely, Low-temperature DRACS Test Facility (LTDF) and High-temperature DRACS Test Facility (HTDF). The LTDF has a nominal power capacity of 6 kW. It uses 1.0-MPa water as the primary coolant, 0.1-MPa water as the secondary coolant, and ambient air as the ultimate heat sink. The main purpose of the LTDF is to examine the couplings among the three natural circulation/convection loops in the DRACS, as well as to provide design and operation experience for the HTDF. An extensive test matrix has

  16. Measures against the adverse impact of natural wind on air-cooled condensers in power plant

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

    The natural wind plays disadvantageous roles in the operation of air-cooled steam condensers in power plant.It is of use to take various measures against the adverse effect of wind for the performance improvement of air-cooled condensers.Based on representative 2×600 MW direct air-cooled power plant,three ways that can arrange and optimize the flow field of cooling air thus enhance the heat transfer of air-cooled condensers were proposed.The physical and mathematical models of air-cooled condensers with various flow leading measures were presented and the flow and temperature fields of cooling air were obtained by CFD simulation.The back pressures of turbine were calculated for different measures on the basis of the heat transfer model of air-cooled condensers.The results show that the performance of air-cooled condensers is improved thus the back pressure of turbine is lowered to some extent by taking measures against the adverse impact of natural wind.

  17. Experimental Analysis Of 80 Tr Capacity Air Cooled Scroll Chiller Using R-22 & R-407c.

    Directory of Open Access Journals (Sweden)

    Mr. Bhikhu B,

    2014-04-01

    Full Text Available In air conditioning systems, chilled water is typically distributed to heat exchangers, or coils, in air handling units or other types of terminal devices which cool the air in their respective space(s, and then the water is recirculated back to the chiller to be cooled again. These cooling coils transfer sensible heat and latent heat from the air to the chilled water, thus cooling and usually dehumidifying the air stream. The experiment works on 80 TR capacity Air Cooled Scroll Chiller systems. Vapour compression refrigeration cycle is used for cooling chilling water. Capacity of compressor is taken same for the different refrigerants used for experimental analysis. During experimental work used R-22 and R-407C as refrigerants. Theoretical COP of system with R- 22 refrigerant is 4.166 and actual COP is 2.227. For 80 TR capacity scroll air cooled chiller, theoretical COP of system with R-407c refrigerant is 3.465 and actual COP is 2.745 respectively. Based on the result analysis the same capacity of air cooling system with scroll compressor Actual COP of R-407C is higher than R-22. It means R-407C is also a alternative refrigerant for air cooled chilling system and also for HVAC system.

  18. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1982

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1983-06-01

    During 1982 the High-Temperature Gas-Cooled Reactor (HTGR) Technology Program at Oak Ridge National Laboratory (ORNL) continued to develop experimental data required for the design and licensing of cogeneration HTGRs. The program involves fuels and materials development (including metals, graphite, ceramic, and concrete materials), HTGR chemistry studies, structural component development and testing, reactor physics and shielding studies, performance testing of the reactor core support structure, and HTGR application and evaluation studies.

  19. A dynamic model of a passively cooled small modular reactor for controller design purposes

    Energy Technology Data Exchange (ETDEWEB)

    Arda, Samet E., E-mail: s.e.arda@asu.edu; Holbert, Keith E., E-mail: holbert@asu.edu

    2015-08-15

    Highlights: • A mathematical dynamic model is developed for a passively cooled small modular reactor. • Reactor response associated single-phase natural circulation is analyzed. • A moving boundary model for a helical-coil steam generator is analyzed. • Dynamic responses of the overall model to representative perturbations are evaluated. • This compact model can be utilized for control system design. - Abstract: An analytical dynamic model for a passively cooled small modular reactor (SMR) is developed using a state-variable lumped parameter approach. Reactor power is represented by the generation time formulation of the point kinetics equations with a single combined neutron precursor group. The heat transfer process in the core is described via an overall heat transfer coefficient by defining two coolant lumps paired to a single fuel lump. In addition, a thermal–hydraulics model for single-phase natural circulation is incorporated. For the helical-coil steam generator, a moving-boundary model including subcooled, two-phase, and superheated regions is utilized. Finally, the hot leg riser and downcomer regions are expressed by first-order lags. The performance of the overall system described by ordinary differential equations (ODEs) is evaluated by the Simulink dynamic environment and directly using a MATLAB ODE solver recommended for stiff systems. Simulation results based on NuScale SMR design data show that the initial steady-state values for 100% power are within range of the design data and the model can predict the system dynamics due to typical perturbations, e.g., control rod movement and change in feedwater mass flow rate and temperature. The model developed in this work can be utilized as a foundation for designing and testing a suitable control algorithm for reactor thermal power.

  20. Enhancement and performance evaluation for heat transfer of air cooling zone for reduction system of sponge titanium

    Science.gov (United States)

    Wang, Wenhao; Wu, Fuzhong; Jin, Huixin

    2016-05-01

    Since the magnesiothermic reduction employed in current sponge titanium is a highly exothermic reaction, the TiCl4 feed rate is carried out slowly to keep a suitable temperature in reduction reactor, which accounts for an extremely low level of productivity and energy efficiency. In order to shorten the production cycle and improve the energy efficiency, an enhancing scheme is proposed to enhance the heat transfer of air cooling zone for reduction system. The air cooling zone and enhancing scheme are firstly introduced. And then, the heat transfer characteristics of cooling zone are obtained by theoretical analysis and experimental date without enhancing scheme. Finally, the enhancement is analyzed and evaluated. The results show that the fitting results of heat transfer coefficients can be used to evaluate the heat transfer enhancement of cooling zone. Heat sources temperatures have a limited decreasing, heat transfer rate increases obviously with the enhanced cooling, and the TiCl4 feed rate can be increased significantly by 9.61 %. And the measured and calculated results are good enough to meet the design requirements.

  1. Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point Design

    Energy Technology Data Exchange (ETDEWEB)

    Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-02-01

    The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include TRISO particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Several preconceptual and conceptual design efforts that have been conducted on FHR concepts bear a significant influence on the FHR DR design. Specific designs include the Oak Ridge National Laboratory (ORNL) advanced high-temperature reactor (AHTR) with 3400/1500 MWt/megawatts of electric output (MWe), as well as a 125 MWt small modular AHTR (SmAHTR) from ORNL. Other important examples are the Mk1 pebble bed FHR (PB-FHR) concept from the University of California, Berkeley (UCB), and an FHR test reactor design developed at the Massachusetts Institute of Technology (MIT). The MIT FHR test reactor is based on a prismatic fuel platform and is directly relevant to the present FHR DR design effort. These FHR concepts are based on reasonable assumptions for credible commercial prototypes. The FHR DR concept also directly benefits from the operating experience of the Molten Salt Reactor Experiment (MSRE), as well as the detailed design efforts for a large molten salt reactor concept and its breeder variant, the Molten Salt Breeder Reactor. The FHR DR technology is most representative of the 3400 MWt AHTR

  2. Test case specifications for coupled neutronics-thermal hydraulics calculation of Gas-cooled Fast Reactor

    Science.gov (United States)

    Osuský, F.; Bahdanovich, R.; Farkas, G.; Haščík, J.; Tikhomirov, G. V.

    2017-01-01

    The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.

  3. Gas-Cooled Fast Reactor: A Historical Overview and Future Outlook

    Directory of Open Access Journals (Sweden)

    W. F. G. van Rooijen

    2009-01-01

    Full Text Available A review is given of developments in the area of Gas-Cooled Fast Reactors (GCFR in the period from roughly 1960 until 1980. During that period, the GCFR concept was expected to increase the breeding gain, the thermal efficiency of a nuclear power plant, and alleviate some of the problems associated with liquid metal coolants. During this period, the GCFR concept was found to be more challenging than liquid-metal-cooled reactors, and none were ever constructed. In the second part of the paper, we provide an overview of the investigations on GCFR since the year 2000, when the Generation IV Initiative rekindled interest in this reactor type. The new GCFR concepts focus primarily on sustainable nuclear power, with very efficient resource use, minimum waste, and a very strong focus on (passive safety. An overview is presented of the main design characteristics of these Gen IV GCFRs, and a literature list is provided to guide the interested reader towards more detailed publications.

  4. Pore Scale Thermal Hydraulics Investigations of Molten Salt Cooled Pebble Bed High Temperature Reactor with BCC and FCC Configurations

    Directory of Open Access Journals (Sweden)

    Shixiong Song

    2014-01-01

    CFD results and empirical correlations’ predictions of pressure drop and local Nusselt numbers. Local pebble surface temperature distributions in several default conditions are investigated. Thermal removal capacities of molten salt are confirmed in the case of nominal condition; the pebble surface temperature under the condition of local power distortion shows the tolerance of pebble in extreme neutron dose exposure. The numerical experiments of local pebble insufficient cooling indicate that in the molten salt cooled pebble bed reactor, the pebble surface temperature is not very sensitive to loss of partial coolant. The methods and results of this paper would be useful for optimum designs and safety analysis of molten salt cooled pebble bed reactors.

  5. Feedback linearization based control of a variable air volume air conditioning system for cooling applications.

    Science.gov (United States)

    Thosar, Archana; Patra, Amit; Bhattacharyya, Souvik

    2008-07-01

    Design of a nonlinear control system for a Variable Air Volume Air Conditioning (VAVAC) plant through feedback linearization is presented in this article. VAVAC systems attempt to reduce building energy consumption while maintaining the primary role of air conditioning. The temperature of the space is maintained at a constant level by establishing a balance between the cooling load generated in the space and the air supply delivered to meet the load. The dynamic model of a VAVAC plant is derived and formulated as a MIMO bilinear system. Feedback linearization is applied for decoupling and linearization of the nonlinear model. Simulation results for a laboratory scale plant are presented to demonstrate the potential of keeping comfort and maintaining energy optimal performance by this methodology. Results obtained with a conventional PI controller and a feedback linearizing controller are compared and the superiority of the proposed approach is clearly established.

  6. Electroremediation of air pollution control residues in a continuous reactor

    DEFF Research Database (Denmark)

    Jensen, Pernille Erland; Ferreira, Célia M. D.; Hansen, Henrik K.

    2010-01-01

    Air pollution control (APC) residue from municipal solid waste incineration is considered hazardous waste due to its alkalinity and high content of salts and mobile heavy metals. Various solutions for the handling of APC-residue exist, however most commercial solutions involve landfilling. A demand...... were made with raw residue, water-washed residue, acid washed residue and acid-treated residue with emphasis on reduction of heavy metal mobility. Main results indicate that the reactor successfully removes toxic elements lead, copper, cadmium and zinc from the feed stream, suggesting...

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

    Directory of Open Access Journals (Sweden)

    Yeong-il Kim

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  9. CFD Analysis of the Fuel Temperature in High Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    In, W. K.; Chun, T. H.; Lee, W. J.; Chang, J. H. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    High temperature gas-cooled reactors (HTGR) have received a renewed interest as potential sources for future energy needs, particularly for a hydrogen production. Among the HTGRs, the pebble bed reactor (PBR) and a prismatic modular reactor (PMR) are considered as the nuclear heat source in Korea's nuclear hydrogen development and demonstration project. PBR uses coated fuel particles embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the core during an operation. PMR uses graphite fuel blocks which contain cylindrical fuel compacts consisting of the fuel particles. The fuel blocks also contain coolant passages and locations for absorber and control material. The maximum fuel temperature in the core hot spot is one of the important design parameters for both PBR and PMR. The objective of this study is to predict the fuel temperature distributions in PBR and PMR using a computational fluid dynamics(CFD) code, CFX-5. The reference reactor designs used in this analysis are PBMR400 and GT-MHR600.

  10. Flow boiling CHF enhancement in an external reactor vessel cooling (ERVC) channel using graphene oxide nanofluid

    Energy Technology Data Exchange (ETDEWEB)

    Park, Seong Dae; Bang, In Cheol, E-mail: icbang@unist.ac.kr

    2013-12-15

    Highlights: • We investigate CHF limits of graphene oxide nanofluid for IVR-ERVC. • Graphene oxide nanofluid enhanced CHF up to about 20%. • CHF enhancement can be explained by the improved thermal activity. - Abstract: External reactor vessel cooling for in-vessel retention of corium is an important concept to mitigate the consequences of a severe accident by flooding the reactor cavity. Although this system has some merits, it is restricted by the capacity of heat removal through the nucleate boiling on the outer surface of the reactor. In this study, the graphene oxide (GO) nanofluid at 0.0001 vol% was used to enhance the critical heat flux (CHF). The CHF tests were conducted with a closed-loop facility. Test section simulated the reactor vessel of APR-1400 with a small scale. The test results show about ∼20% enhancement of CHF at 50 and 100 kg/m{sup 2} s under a 10 K subcooling condition. It means that the additional thermal margin could be acquired by just adding the GO nanoparticles to the flooding water without severe economic concerns. It is also found that this CHF enhancement is caused by coating the graphene oxide nanoparticles on the heated surface. However, the sessile drop tests on the coated heater surface show that the wettability of GO coated surface is not improved. The results of IR thermography show that one of the promising reasons is the change of thermal activity due to the coated GO nanoparticles on the heated surface.

  11. Comparative analysis of thorium and uranium fuel for transuranic recycle in a sodium cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    C. Fiorina; N. E. Stauff; F. Franceschini; M. T. Wenner; A. Stanculescu; T. K. Kim; A. Cammi; M. E. Ricotti; R. N. Hill; T. A. Taiwo; M. Salvatores

    2013-12-01

    The present paper compares the reactor physics and transmutation performance of sodium-cooled Fast Reactors (FRs) for TRansUranic (TRU) burning with thorium (Th) or uranium (U) as fertile materials. The 1000 MWt Toshiba-Westinghouse Advanced Recycling Reactor (ARR) conceptual core has been used as benchmark for the comparison. Both burner and breakeven configurations sustained or started with a TRU supply, and assuming full actinide homogeneous recycle strategy, have been developed. State-of-the-art core physics tools have been employed to establish fuel inventory and reactor physics performances for equilibrium and transition cycles. Results show that Th fosters large improvements in the reactivity coefficients associated with coolant expansion and voiding, which enhances safety margins and, for a burner design, can be traded for maximizing the TRU burning rate. A trade-off of Th compared to U is the significantly larger fuel inventory required to achieve a breakeven design, which entails additional blankets at the detriment of core compactness as well as fuel manufacturing and separation requirements. The gamma field generated by the progeny of U-232 in the U bred from Th challenges fuel handling and manufacturing, but in case of full recycle, the high contents of Am and Cm in the transmutation fuel impose remote fuel operations regardless of the presence of U-232.

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

    Energy Technology Data Exchange (ETDEWEB)

    Almeida, Lucilla C.; Su, Jian, E-mail: lucillalmeida@gmail.com, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Aguirre, Joao, E-mail: aguirre@rocky-dem.com [Engineering Simulation and Scientific Software (ESSS), Rio de Janeiro, RJ (Brazil)

    2015-07-01

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

  13. HYBRID SULFUR CYCLE FLOWSHEETS FOR HYDROGEN PRODUCTION USING HIGH-TEMPERATURE GAS-COOLED REACTORS

    Energy Technology Data Exchange (ETDEWEB)

    Gorensek, M.

    2011-07-06

    Two hybrid sulfur (HyS) cycle process flowsheets intended for use with high-temperature gas-cooled reactors (HTGRs) are presented. The flowsheets were developed for the Next Generation Nuclear Plant (NGNP) program, and couple a proton exchange membrane (PEM) electrolyzer for the SO2-depolarized electrolysis step with a silicon carbide bayonet reactor for the high-temperature decomposition step. One presumes an HTGR reactor outlet temperature (ROT) of 950 C, the other 750 C. Performance was improved (over earlier flowsheets) by assuming that use of a more acid-tolerant PEM, like acid-doped poly[2,2'-(m-phenylene)-5,5'-bibenzimidazole] (PBI), instead of Nafion{reg_sign}, would allow higher anolyte acid concentrations. Lower ROT was accommodated by adding a direct contact exchange/quench column upstream from the bayonet reactor and dropping the decomposition pressure. Aspen Plus was used to develop material and energy balances. A net thermal efficiency of 44.0% to 47.6%, higher heating value basis is projected for the 950 C case, dropping to 39.9% for the 750 C case.

  14. Fluoride-Salt-Cooled High-Temperature Reactor (FHR) for Power and Process Heat

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, Charles [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Hu, Lin-wen [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Peterson, Per [Univ. of California, Berkeley, CA (United States); Sridharan, Kumar [Univ. of Wisconsin, Madison, WI (United States)

    2015-01-21

    In 2011 the U.S. Department of Energy through its Nuclear Energy University Program (NEUP) awarded a 3- year integrated research project (IRP) to the Massachusetts Institute of Technology (MIT) and its partners at the University of California at Berkeley (UCB) and the University of Wisconsin at Madison (UW). The IRP included Westinghouse Electric Company and an advisory panel chaired by Regis Matzie that provided advice as the project progressed. The first sentence of the proposal stated the goals: The objective of this Integrated Research Project (IRP) is to develop a path forward to a commercially viable salt-cooled solid-fuel high-temperature reactor with superior economic, safety, waste, nonproliferation, and physical security characteristics compared to light-water reactors. This report summarizes major results of this research.

  15. A small high temperature gas cooled reactor for nuclear marine propulsion

    Energy Technology Data Exchange (ETDEWEB)

    Brugiere, F.; Sillon, C. [Ecole des Applications Militaires de l' Energie Atomique, 50 - Cherbourg (France); Foster, A.; Hamilton, P.; Jewer, S.; Thompson, A.C. [Defence College of Electromechanical Engineering, Nuclear Dept., Military Rd, Gosport (United Kingdom); Kingston, T.; Williams, A.M.; Beeley, P.A. [Rolls-Royce (Marine Power), Raynesway, Derby (United Kingdom)

    2007-07-01

    Results from a design study for a hypothetical nuclear marine propulsion plant are presented. The plant utilizes a small High Temperature Gas Cooled Reactor (HTGCR) similar to the GTHTR300 design by the Japan Atomic Energy Agency with power being generated by a direct cycle gas turbine. The GTHTR300 design is modified in order to achieve the required power of 80 MWth and core lifetime of approximately 10 years. Thermal hydraulic analysis shows that in the event of a complete loss of flow accident the hot channel fuel temperature exceeds the 1600 Celsius degrees limit due to the high power peaking in assemblies adjacent to the inner reflector. Reactor dynamics shows oscillatory behaviour in rapid power transients. An automatic control rod system is suggested to overcome this problem. (authors)

  16. Development of objective provision trees for Sodium-Cooled Fast Reactor Defense-in-depth evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Huichang [TUEV Rheinland Korea Ltd., Seoul (Korea, Republic of); Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-05-15

    KALIMER is one of sodium-cooled fast reactor and being developed by Korea Atomic Energy Research Institute (KAERI), was developed and suggested in this paper. Developed OPT is for the defense-in-depth level 3, core heat removal safety function. Using OPT method, the evaluation of defense-in-depth implementation for the design features of KALIMER reactors were tried in this study. To utilize the design information of KALIMER, challenges in OPTs which are under development in this study, were identified based on the system physical boundaries. This approach make the identification of possible and postulated challenges much clear and this will be a benefit to further identification of provisions in KALIMER design. OPTs for other levels of defense-in-depth and other safety functions are under development.

  17. High temperature gas-cooled reactor (HTGR) graphite pebble fuel: Review of technologies for reprocessing

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-08

    This report reviews literature on reprocessing high temperature gas-cooled reactor graphite fuel components. A basic review of the various fuel components used in the pebble bed type reactors is provided along with a survey of synthesis methods for the fabrication of the fuel components. Several disposal options are considered for the graphite pebble fuel elements including the storage of intact pebbles, volume reduction by separating the graphite from fuel kernels, and complete processing of the pebbles for waste storage. Existing methods for graphite removal are presented and generally consist of mechanical separation techniques such as crushing and grinding chemical techniques through the use of acid digestion and oxidation. Potential methods for reprocessing the graphite pebbles include improvements to existing methods and novel technologies that have not previously been investigated for nuclear graphite waste applications. The best overall method will be dependent on the desired final waste form and needs to factor in the technical efficiency, political concerns, cost, and implementation.

  18. Power flattening on modified CANDLE small long life gas-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Monado, Fiber [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung, Indonesia and Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Su' ud, Zaki; Waris, Abdul; Basar, Khairul [Nuclear Physics and Biophysics Research Group, Dept. of Physics, Faculty of Mathematics and Natural Sciences, Bandung Institute of Technology, Bandung (Indonesia); Ariani, Menik [Dept. of Physics, Faculty of Mathematics and Natural Sciences, Sriwijaya University (Indonesia); Sekimoto, Hiroshi [CRINES, Tokyo Institute of Technology, O-okoyama, Meguro-ku, Tokyo 152-8550 (Japan)

    2014-09-30

    Gas-cooled Fast Reactor (GFR) is one of the candidates of next generation Nuclear Power Plants (NPPs) that expected to be operated commercially after 2030. In this research conceptual design study of long life 350 MWt GFR with natural uranium metallic fuel as fuel cycle input has been performed. Modified CANDLE burn-up strategy with first and second regions located near the last region (type B) has been applied. This reactor can be operated for 10 years without refuelling and fuel shuffling. Power peaking reduction is conducted by arranging the core radial direction into three regions with respectively uses fuel volume fraction 62.5%, 64% and 67.5%. The average power density in the modified core is about 82 Watt/cc and the power peaking factor decreased from 4.03 to 3.43.

  19. Comparison of containment systems for large sodium-cooled breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Seeman, S.E.; Armstrong, G.R.

    1978-04-01

    Three types of containment configurations for large breeder reactors (1000 MWe/2500 MWt) were compared to determine characteristic responses of each type to selected hypothetical accident conditions. The three types of system configurations were the single, double and containment-confinement designs. Two types of postulated hypothetical accidents were applied as a basis for evaluation--the energetic core disassambly accident and the whole core melt-through accident. More than 20 separate cases were analyzed in which design and accident assumptions were varied. Based on this study the containment-confinement configuration is shown to be the most promising for limiting off-site radiation doses for large sodium-cooled breeder reactors given the specific assumptions on the initial accident conditions.

  20. Overall system description and safety characteristics of Prototype Gen IV Sodium Cooled Fast Reactor in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Jae Woon; Chang, Jin Wook; Lim, Jae Yong; Cheon, Jin Sik; Lee, Tae Ho; Kim, Sung Kyun; Lee, Kwi Lim; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The Prototype Gen IV sodium cooled fast reactor (PGSFR) has been developed for the last 4 years, fulfilling the technology demonstration of the burning capability of transuranic elements included in light water reactor spent nuclear fuel. The PGSFR design has been focused on the robustness of safety systems by enhancing inherent safety characteristics of metal fuel and strengthening passive safety features using natural circulation and thermal expansion. The preliminary safety information document as a major outcome of the first design phase of PGSFR development was issued at the end of 2015. The project entered the second design phase at the beginning of 2016. This paper summarizes the overall structures, systems, and components of nuclear steam supply system and safety characteristics of the PGSFR. The research and development activities to demonstrate the safety performance are also briefly introduced in the paper.

  1. Minor actinides impact on basic safety parameters of medium-sized sodium-cooled fast reactor

    Directory of Open Access Journals (Sweden)

    Darnowski Piotr

    2015-03-01

    Full Text Available An analysis of the influence of addition of minor actinides (MA to the fast reactor fuel on the most important safety characteristics was performed. A special emphasis was given to the total control rods worth in order to describe qualitatively and quantitatively its change with MA content. All computations were performed with a homogeneous assembly model of modified BN-600 sodium-cooled fast reactor core with 0, 3 and 6% of MA. A model was prepared for the Monte Carlo neutron transport code MCNP5 for fresh fuel in the beginning-of-life (BOL state. Additionally, some other parameters, such as Doppler constant, sodium void reactivity, delayed neutron fraction, neutron fluxes and neutron spectra distribution, were computed and their change with MA content was investigated. Study indicates that the total control rods worth (CRW decreases with increasing MA inventory in the fuel and confirms that the addition of MA has a negative effect on the delayed neutron fraction.

  2. A water cooled, lithium lead breeding blanket for a DEMO fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Casini, G.; Rieger, M.; Biggio, M.; Farfaletti-Casali, F.; Tominetti, S.; Wu, J.; Zucchetti, M. (Commission of the European Communities, Ispra (Italy). Joint Research Centre); Labbe, P.; Baraer, L.; Gervaise, G.; Giancarli, L.; Roze, M.; Severi, Y.; Quintric-Bossy, J. (CEA Centre d' Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France))

    1991-04-01

    The main features of a tritium breeding blanket for a Demonstration Power Reactor involving the eutectic Pb-17Li as liquid breeder and water as coolant are presented. The configuration of the blanket segments and breeder modules as well as their arrangement inside the reactor vacuum vessel are outlined. The main design aspects and the corresponding design limits are reviewed, namely those related to thermomechanics, neutronics, magneto-hydrodynamics, tritium permeation and recovery. First results of safety analysis, in particular those connected with the rupture of a coolant tube in the breeder module are presented and discussed. As a conclusion, the feasibility of the concept look attractive. A problem which requires further investigation is that of the tritium self-sufficiency. It is shown that a net tritium production near to one can be obtained if berylium tiles are placed in front of the plasma, provided that they are cooled by heavy water. (orig.).

  3. 2400MWt GAS-COOLED FAST REACTOR DHR STUDIES STATUS UPDATE.

    Energy Technology Data Exchange (ETDEWEB)

    CHENG,L.Y.; LUDEWIG, H.

    2007-06-01

    A topical report on demonstrating the efficacy of a proposed hybrid active/passive combination approach to the decay heat removal for an advanced 2400MWt GEN-IV gas-cooled fast reactor was published in March 2006. The analysis was performed with the system code RELAP5-3D (version 2.4.1.1a) and the model included the full complement of the power conversion unit (PCU): heat exchange components (recuperator, precooler, intercooler) and rotating machines (turbine, compressor). A re-analysis of the success case in Ref is presented in this report. The case was redone to correct unexpected changes in core heat structure temperatures when the PCU model was first integrated with the reactor model as documented in Ref [1]. Additional information on the modeling of the power conversion unit and the layout of the heat exchange components is provided in Appendix A.

  4. Corrosion mechanisms of candidate structural materials for supercritical water-cooled reactor

    Institute of Scientific and Technical Information of China (English)

    Lefu ZHANG; Fawen ZHU; Rui TANG

    2009-01-01

    Nickel-based alloys, austenitic stainless steel, ferritic/martensitic heat-resistant steels, and oxide dispersion strengthened steel are presently considered to be the candidate structural or fuel-cladding materials for supercritical water-cooled reactor (SCWR), one of the promising generation IV reactor for large-scale electric power production. However, corrosion and stress corrosion cracking of these candidate alloys still remain to be a major problem in the selection of nuclear fuel cladding and other structural materials, such as water rod. Survey of literature and experimental results reveal that the general corrosion mechanism of those candidate materials exhibits quite complicated mechanism in high-temperature and high-pressure supercritical water. Formation of a stable protective oxide film is the key to the best corrosion-resistant alloys. This paper focuses on the mechanism of corrosion oxide film breakdown for SCWR candidate materials.

  5. A Conceptual Study of a Supercritical CO2-Cooled Micro Modular Reactor

    Directory of Open Access Journals (Sweden)

    Hwanyeal Yu

    2015-12-01

    Full Text Available A neutronics conceptual study of a supercritical CO2-cooled micro modular reactor (MMR has been performed in this work. The suggested MMR is an extremely compact and truck-transportable nuclear reactor. The thermal power of the MMR is 36.2 MWth and it is designed to have a 20-year lifetime without refueling. A salient feature of the MMR is that all the components including the generator are integrated in a small reactor vessel. For a minimal volume and long lifetime of the MMR core, a fast neutron spectrum is utilized in this work. To enhance neutron economy and maximize the fuel volume fraction in the core, a high-density uranium mono-nitride U15N fuel is used in the fast-spectrum MMR. Unlike the conventional supercritical CO2-cooled fast reactors, a replaceable fixed absorber (RFA is introduced in a unique way to minimize the excess reactivity and the power peaking factor of the core. For a compact core design, the drum-type control absorber is adopted as the primary reactivity control mechanism. In this study, the neutronics analyses and depletions have been performed by using the continuous energy Monte Carlo Serpent code with the evaluated nuclear data file ENDF/B-VII.1 Library. The MMR core is characterized in view of several important safety parameters such as control system worth, fuel temperature coefficient (FTC and coolant void reactivity (CVR, etc. In addition, a preliminary thermal-hydraulic analysis has also been performed for the hottest channel of the Korea Advanced Institute of Science and Technology (KAIST MMR.

  6. Sustainability of thorium-uranium in pebble-bed fluoride salt-cooled high temperature reactor

    OpenAIRE

    Zhu Guifeng; Zou Yang; Xu Hongjie

    2016-01-01

    Sustainability of thorium fuel in a Pebble-Bed Fluoride salt-cooled High temperature Reactor (PB-FHR) is investigated to find the feasible region of high discharge burnup and negative Flibe (2LiF-BeF2) salt Temperature Reactivity Coefficient (TRC). Dispersion fuel or pellet fuel with SiC cladding and SiC matrix is used to replace the tristructural-isotropic (TRISO) coated particle system for increasing fuel loading and decreasing excessive moderation. To analyze the neutronic characteristics,...

  7. High Temperature Gas-cooled Reactor Projected Markets and Scoping Economics

    Energy Technology Data Exchange (ETDEWEB)

    Larry Demick

    2010-08-01

    The NGNP Project has the objective of developing the high temperature gas-cooled reactor (HTGR) technology to supply high temperature process heat to industrial processes as a substitute for burning of fossil fuels, such as natural gas. Applications of the HTGR technology that have been evaluated by the NGNP Project for supply of process heat include supply of electricity, steam and high-temperature gas to a wide range of industrial processes, and production of hydrogen and oxygen for use in petrochemical, refining, coal to liquid fuels, chemical, and fertilizer plants.

  8. Qualification of Simulation Software for Safety Assessment of Sodium Cooled Fast Reactors. Requirements and Recommendations

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pointer, William David [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Sieger, Matt [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Flanagan, George F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Moe, Wayne [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); HolbrookINL, Mark [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-04-01

    The goal of this review is to enable application of codes or software packages for safety assessment of advanced sodium-cooled fast reactor (SFR) designs. To address near-term programmatic needs, the authors have focused on two objectives. First, the authors have focused on identification of requirements for software QA that must be satisfied to enable the application of software to future safety analyses. Second, the authors have collected best practices applied by other code development teams to minimize cost and time of initial code qualification activities and to recommend a path to the stated goal.

  9. Metal Fuel Development and Verification for Prototype Generation IV Sodium-Cooled Fast Reactor

    OpenAIRE

    Chan Bock Lee; Jin Sik Cheon; Sung Ho Kim; Jeong-Yong Park; Hyung-Kook Joo

    2016-01-01

    Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U–Zr fuel is a driver for the initial core of the PGSFR, and U–transuranics (TRU)–Zr fuel will gradually replace U–Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U–Zr fuel, work on U–Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U–TRU–Zr fuel uses TRU recovered through pyroelectrochem...

  10. SPARC fast reactor design : Design of two passively safe metal-fuelled sodium-cooled pool-type small modular fast reactors with Autonomous Reactivity Control

    OpenAIRE

    Lindström, Tobias

    2015-01-01

    In this master thesis a small modular sodium-cooled metal-fuelled pool-type fast reactor design, called SPARC - Safe and Passive with Autonomous Reactivity control, has been designed. The long term reactivity changes in the SPARC are managed by implementation of the the Autonomous Reactivity Control (ARC) system, which is the novelty of the design. The overall design is mainly based on the Integral Fast Reactor project (IFR), which experimentally demonstrated the passive safety characteristic...

  11. Membrane-based air conditioning. Decentral desiccant cooling; Membranbasierte Raumklimatisierung. Dezentrale sorptionsgestuetzte Klimatisierung

    Energy Technology Data Exchange (ETDEWEB)

    Rosenbaum, Hannes [Institut fuer Luft- und Kaeltetechnik gGmbH, Dresden (Germany)

    2012-04-15

    Semipermeable membranes have a considerable potential use in air conditioning systems due to their properties. As water resistant but water vapor permeable functional separation layers, they allow desiccant air-conditioning processes in decentral air conditioning systems and a condensate-free air dehumidification at cooling ceiling elements. (orig.)

  12. ANALYSIS OF A HIGH TEMPERATURE GAS-COOLED REACTOR POWERED HIGH TEMPERATURE ELECTROLYSIS HYDROGEN PLANT

    Energy Technology Data Exchange (ETDEWEB)

    M. G. McKellar; E. A. Harvego; A. M. Gandrik

    2010-11-01

    An updated reference design for a commercial-scale high-temperature electrolysis (HTE) plant for hydrogen production has been developed. The HTE plant is powered by a high-temperature gas-cooled reactor (HTGR) whose configuration and operating conditions are based on the latest design parameters planned for the Next Generation Nuclear Plant (NGNP). The current HTGR reference design specifies a reactor power of 600 MWt, with a primary system pressure of 7.0 MPa, and reactor inlet and outlet fluid temperatures of 322°C and 750°C, respectively. The reactor heat is used to produce heat and electric power to the HTE plant. A Rankine steam cycle with a power conversion efficiency of 44.4% was used to provide the electric power. The electrolysis unit used to produce hydrogen includes 1.1 million cells with a per-cell active area of 225 cm2. The reference hydrogen production plant operates at a system pressure of 5.0 MPa, and utilizes a steam-sweep system to remove the excess oxygen that is evolved on the anode (oxygen) side of the electrolyzer. The overall system thermal-to-hydrogen production efficiency (based on the higher heating value of the produced hydrogen) is 42.8% at a hydrogen production rate of 1.85 kg/s (66 million SCFD) and an oxygen production rate of 14.6 kg/s (33 million SCFD). An economic analysis of this plant was performed with realistic financial and cost estimating The results of the economic analysis demonstrated that the HTE hydrogen production plant driven by a high-temperature helium-cooled nuclear power plant can deliver hydrogen at a competitive cost. A cost of $3.03/kg of hydrogen was calculated assuming an internal rate of return of 10% and a debt to equity ratio of 80%/20% for a reactor cost of $2000/kWt and $2.41/kg of hydrogen for a reactor cost of $1400/kWt.

  13. Neutronic assessment of liquid-metal cooled fast reactors using thorium fuel

    Energy Technology Data Exchange (ETDEWEB)

    Pilarski, Stevan [Electricite de France R et D, 1 Avenue du General de Gaulle, 92141 Clamart (France); Institut de Physique Nucleaire d' Orsay, 15 rue Georges Clemenceau 91406 Orsay (France)

    2009-06-15

    The long-term sustainability of atomic fission energy will require the development of new types of reactors, able to exceed the limits of the existing ones in terms of optimal use of natural resources, which clearly necessitates breeding of fissile material. In this context, fast reactors using uranium-plutonium fuel are the most mature solution from an industrial viewpoint. In addition to the obvious interest in terms of fuel resources, there is a major incentive to consider the use of the {sup 232}Th- {sup 233}U fuel cycle as an alternative to the traditional {sup 238}U-{sup 239}Pu cycle for fast reactors: it is an effective way of addressing the safety issue of the highly positive void reactivity effect, which is a well-known problem for liquid-metal cooled fast reactors of commercial size [1]. This work investigates the performance of liquid-metal cooled fast reactors in {sup 232}Th-{sup 233}U fuel cycle and draws a comparison with the traditional {sup 238}U-{sup 239}Pu cycle. Four coolants have been considered: Na, Pb, Mg(17%at.)-Pb and Li(17%at.)-Pb; a simulation of their use in cores ranging from 700 MWth to 3600 MWth has been performed in two-dimensional diffusion theory using the European system of codes ERANOS [2,3] developed at CEA. The performance parameters such as the breeding ratio have been computed for each concept, alongside safety-related parameters: the delayed neutron fraction, the cycle reactivity swing, the Doppler constant and other thermal feedbacks. More specifically, the issue of void reactivity is studied in detail using perturbation theory. These calculations are performed at equilibrium fuel composition and are complemented by the study of the initial fuel loading at start-up which is a mixture of {sup 232}Th-{sup 239}Pu. The isotopic composition of the fissile corresponds to the plutonium available from French reactors in 2035. The conclusions of this work are that near-zero to large negative void reactivity effects can be achieved in

  14. Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Ferroni, Paolo [Westinghouse Electric Company LLC, Cranberry Township, PA (United States). Global Technology Development; Tatli, Emre [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Czerniak, Luke [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Chien, Hual-Te [Argonne National Lab. (ANL), Argonne, IL (United States); Yoichi, Momozaki [Argonne National Lab. (ANL), Argonne, IL (United States); Bakhtiari, Sasan [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-06-29

    The project “Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems” was conducted jointly by Westinghouse Electric Company (Westinghouse) and Argonne National Laboratory (ANL), over the period October 1, 2013- March 31, 2016. The project’s motivation was the need to provide designers of Sodium Fast Reactors (SFRs) with a validated, state-of-the-art computational tool for the prediction of sodium oxide (Na2O) deposition in small-diameter sodium heat exchanger (HX) channels, such as those in the diffusion bonded HXs proposed for SFRs coupled with a supercritical CO2 (sCO2) Brayton cycle power conversion system. In SFRs, Na2O deposition can potentially occur following accidental air ingress in the intermediate heat transport system (IHTS) sodium and simultaneous failure of the IHTS sodium cold trap. In this scenario, oxygen can travel through the IHTS loop and reach the coldest regions, represented by the cold end of the sodium channels of the HXs, where Na2O precipitation may initiate and continue. In addition to deteriorating HX heat transfer and pressure drop performance, Na2O deposition can lead to channel plugging especially when the size of the sodium channels is small, which is the case for diffusion bonded HXs whose sodium channel hydraulic diameter is generally below 5 mm. Sodium oxide melts at a high temperature well above the sodium melting temperature such that removal of a solid plug such as through dissolution by pure sodium could take a lengthy time. The Sodium Plugging Phenomena Loop (SPPL) was developed at ANL, prior to this project, for investigating Na2O deposition phenomena within sodium channels that are prototypical of the diffusion bonded HX channels envisioned for SFR-sCO2 systems. In this project, a Computational Fluid Dynamic (CFD) model capable of simulating the thermal-hydraulics of the SPPL test

  15. Citywide Impacts of Cool Roof and Rooftop Solar Photovoltaic Deployment on Near-Surface Air Temperature and Cooling Energy Demand

    Science.gov (United States)

    Salamanca, F.; Georgescu, M.; Mahalov, A.; Moustaoui, M.; Martilli, A.

    2016-10-01

    Assessment of mitigation strategies that combat global warming, urban heat islands (UHIs), and urban energy demand can be crucial for urban planners and energy providers, especially for hot, semi-arid urban environments where summertime cooling demands are excessive. Within this context, summertime regional impacts of cool roof and rooftop solar photovoltaic deployment on near-surface air temperature and cooling energy demand are examined for the two major USA cities of Arizona: Phoenix and Tucson. A detailed physics-based parametrization of solar photovoltaic panels is developed and implemented in a multilayer building energy model that is fully coupled to the Weather Research and Forecasting mesoscale numerical model. We conduct a suite of sensitivity experiments (with different coverage rates of cool roof and rooftop solar photovoltaic deployment) for a 10-day clear-sky extreme heat period over the Phoenix and Tucson metropolitan areas at high spatial resolution (1-km horizontal grid spacing). Results show that deployment of cool roofs and rooftop solar photovoltaic panels reduce near-surface air temperature across the diurnal cycle and decrease daily citywide cooling energy demand. During the day, cool roofs are more effective at cooling than rooftop solar photovoltaic systems, but during the night, solar panels are more efficient at reducing the UHI effect. For the maximum coverage rate deployment, cool roofs reduced daily citywide cooling energy demand by 13-14 %, while rooftop solar photovoltaic panels by 8-11 % (without considering the additional savings derived from their electricity production). The results presented here demonstrate that deployment of both roofing technologies have multiple benefits for the urban environment, while solar photovoltaic panels add additional value because they reduce the dependence on fossil fuel consumption for electricity generation.

  16. Control of cooling processes with forced-air aimed at efficiency energetic and the cooling time of horticultural products

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Joao Carlos Teles Ribeiro da; Mederos, Barbara Janet Teruel [Universidade Estadual de Campinas (FEAGRI/UNICAMP), SP (Brazil). Fac. de Engenharia Agricola

    2008-07-01

    The application of cooling technologies for the conservation of horticultural products is one of the stages the Cold Chain. In Brazil particularly, as a country with tropical climate with average high temperature almost all year, the application of these technologies is very important because the shelf-life of fresh horticultural products, with quality that the market demands, is directly related to temperature. In particular, the systems of forced air cooling operate according to the flow of air predetermined in the project according to the quantity of product to cool. When actual conditions differ from considerations of the project, as to the quantity of product, a situation very common in agricultural properties and packing houses, the fan will continue providing the nominal flow rate, causing alteration of the cost-benefit relation of process. This project aims at the development of a micro-processing equipment (output current of 4 to 20 mA) to control the rotational speed of the motor of the fan systems, air forced through an inverter of frequency. The objective is development of a Man-Machine Interface, based on an algorithm, which, through the introduction of mass product data and the automatic acquisition of data from temperature of the product and the camera, is calculated the cooling time. The rotation of the engine fan will be amended automatically, to maintain air flow with a proper cost-benefit, in connection with the reduction of cooling time, energy consumption, for the increasing the shelf life of products. (author)

  17. Study on the properties of the fuel compact for High Temperature Gas-cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chung-yong; Lee, Sung-yong; Choi, Min-young; Lee, Seung-jae; Jo, Young-ho [KEPCO Nuclear Fuel, Daejeon (Korea, Republic of); Lee, Young-woo; Cho, Moon-sung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    High Temperature Gas-cooled Reactors (HTGR), one of the Gen-IV reactors, have been using the fuel element which is manufactured by the graphite matrix, surrounding Tristructural-isotropic (TRISO)-coated Uranium particles. Factors with these characteristics effecting on the matrix of fuel compact are chosen and their impacts on the properties are studied. The fuel elements are considered with two types of concepts for HTGR, which are the block type reactor and the pebble bed reactor. In this paper, the cylinder-formed fuel element for the block type reactor is focused on, which consists of the large part of graphite matrix. One of the most important properties of the graphite matrix is the mechanical strength with the high reliability because the graphite matrix should be enabled to protect the TRISO particles from the irradiation environment and the impact from the outside. In this study, the three kinds of candidate graphites and the two kinds of candidate binder (Phenol and Polyvinyl butyral) were chosen and mixed with each other, formed and heated to measure mechanical properties. The objective of this research is to optimize the materials and composition of the mixture and the forming process by evaluating the mechanical properties before/after carbonization and heat treatment. From the mechanical test results, the mechanical properties of graphite pellets was related to the various conditions such as the contents and kinds of binder, the kinds of graphite and the heat treatments. In the result of the compressive strength and Vicker's hardness, the 10 wt% phenol binder added R+S graphite pellet was relatively higher mechanical properties than other pellets. The contents of Phenol binder, the kinds of graphite powder and the temperature of carbonization and heat treatment are considered important factors for the properties. To optimize the mechanical properties of fuel elements, the role of binders and the properties of graphites will be investigated as

  18. An experimental study of hypervapotron structure in external reactor vessel cooling

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Yufeng; Zhang, Ming [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Hou, Fangxin [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing (China); Gao, Tianfang [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Chen, Peipei, E-mail: chenpeipei@snptc.com.cn [State Power Investment Group Corporation, Beijing (China)

    2016-07-15

    Highlights: • Experiments are performed to study the application of hypervapotron in ERVC design. • CHF experiments on two surfaces are conducted under different flow conditions. • Hypervapotron improves CHF performance by 40–60% compared with smooth surface. • Visualization shows fin structure removes vapor mushroom for better liquid supply. - Abstract: In vessel retention (IVR) is one of the key strategies for many advanced LWR designs to mitigate postulated severe accidents. The success of IVR substantially relies on external reactor vessel cooling (ERVC) by which the decay heat is removed from the melt core in the reactor vessel lower head. The main challenge of IVR is to provide an adequate safety margin of ERVC against critical heat flux (CHF) of subcooled flow boiling in the reactor lower head flow channel. Due to uncertainties in corium melt pool configuration, large CHF margin of ERVC is usually required by regulatory authorities to demonstrate reliability of severe accident mitigation methods. Various CHF enhancement designs have been proposed and studied in literature. In this paper, an experimental study of hypervapotron structure as a novel design to improve CHF performance of ERVC is conducted. Hypervapotron is chosen as one of the potential engineering options for International Thermonuclear Experimental Reactor (ITER) program as a divertor structure to remove highly intense heat from fusion chamber. This study is to conduct CHF experiments at typical PWR ERVC working conditions. The CHF experiments are performed in a 30 mm by 61 mm rectangular flow channel with a 200 mm long heated surface along the flow direction. Both smooth and hypervapotron surface are tested at various inclination angles of the test section to simulate various positions of the reactor lower head. The hypervapotron is found to have a 40–60% CHF improvement compared with the smooth surface. The high speed visualization indicates that hypervapotron is able to

  19. Simulation and Optimization of Air-Cooled PEMFC Stack for Lightweight Hybrid Vehicle Application

    Directory of Open Access Journals (Sweden)

    Jingming Liang

    2015-01-01

    Full Text Available A model of 2 kW air-cooled proton exchange membrane fuel cell (PEMFC stack has been built based upon the application of lightweight hybrid vehicle after analyzing the characteristics of heat transfer of the air-cooled stack. Different dissipating models of the air-cooled stack have been simulated and an optimal simulation model for air-cooled stack called convection heat transfer (CHT model has been figured out by applying the computational fluid dynamics (CFD software, based on which, the structure of the air-cooled stack has been optimized by adding irregular cooling fins at the end of the stack. According to the simulation result, the temperature of the stack has been equally distributed, reducing the cooling density and saving energy. Finally, the 2 kW hydrogen-air air-cooled PEMFC stack is manufactured and tested by comparing the simulation data which is to find out its operating regulations in order to further optimize its structure.

  20. Cooling Performance Characteristics on Mobile Air-Conditioning System for Hybrid Electric Vehicles

    Directory of Open Access Journals (Sweden)

    Ho-Seong Lee

    2013-01-01

    Full Text Available This study investigates the cooling performance characteristics of the mobile air-conditioning system using R744 (CO2 for the hybrid electric vehicle as an alternative to both the R-134a and the conventional air-conditioning system. The developed air-conditioning system is operated with an electric driven compressor in the battery driving mode and a belt driven compressor in the engine driving mode. The cooling performance characteristics of the developed system have been analyzed by experiments under various operating conditions of inlet air temperature, air flow rates for the gas cooler side and evaporator side, and electric compressor revolution respectively. As a result, cooling performances of the tested air-conditioning system for the EDC driving mode (electricity driven compressor were better than those for the BDC driving mode (belt driven compressor. The cooling capacity and cooling COP of the tested air-conditioning system for both driving modes were over 5.0 kW and 2.0, respectively. The observed cooling performance of the tested air-conditioning system may be sufficient for the cabin cooling of hybrid electric vehicles.

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

    Science.gov (United States)

    Ilham, Muhammad; Su’ud, Zaki

    2017-01-01

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

  2. Piping Flexibility Analysis of the Primary Cooling System of TRIGA 2000 Bandung Reactor due to Earthquake

    Directory of Open Access Journals (Sweden)

    H.P. Rahardjo

    2011-08-01

    Full Text Available Earthquakes in a nuclear installation can overload a piping system which is not flexible enough. These loads can be forces, moments and stresses working on the pipes or equipments. If the load is too large and exceed the allowable limits, the piping and equipment can be damaged and lead to overall system operation failure. The load received by piping systems can be reduced by making adequate piping flexibility, so all the loads can be transmitted homogenously throughout the pipe without load concentration at certain point. In this research the analysis of piping stress has been conducted to determine the size of loads that occured in the piping of primary cooling system of TRIGA 2000 Reactor, Bandung if an earthquake happened in the reactor site. The analysis was performed using Caesar II software-based finite element method. The ASME code B31.1 arranging the design of piping systems for power generating system (Power Piping Code was used as reference analysis method. Modeling of piping systems was based on the cooling piping that has already been installed and the existing data reported in Safety Analysis Reports (SARs of TRIGA 2000 reactor, Bandung. The quake considered in this analysis is the earthquake that occurred due to the Lembang fault, since it has the Peak Ground Acceleration (PGA in the Bandung TRIGA 2000 reactor site. The analysis results showed that in the static condition for sustain and expansion loads, the stress fraction in all piping lines does not exceed the allowable limit. However, during operation moment, in dynamic condition, the primary cooling system is less flexible at sustain load, ekspansi load, and combination load and the stress fraction have reached 95,5%. Therefore a pipeline modification (rerouting is needed to make pipe stress does not exceed the allowable stress. The pipeline modification was carried out by applied a gap of 3 mm in the X direction of the support at node 25 and eliminate the support at the node

  3. Cooling Performance Characteristics on Mobile Air-Conditioning System for Hybrid Electric Vehicles

    OpenAIRE

    Ho-Seong Lee; Moo-Yeon Lee

    2013-01-01

    This study investigates the cooling performance characteristics of the mobile air-conditioning system using R744 (CO2) for the hybrid electric vehicle as an alternative to both the R-134a and the conventional air-conditioning system. The developed air-conditioning system is operated with an electric driven compressor in the battery driving mode and a belt driven compressor in the engine driving mode. The cooling performance characteristics of the developed system have been analyzed by experim...

  4. Contributions to the neutronic analysis of a gas-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Martin-del-Campo, Cecilia, E-mail: cecilia.martin.del.campo@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Reyes-Ramirez, Ricardo, E-mail: ricarera@yahoo.com.mx [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Francois, Juan-Luis, E-mail: juan.luis.francois@gmail.com [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico); Reinking-Cejudo, Arturo G., E-mail: reinking@servidor.unam.mx [Departamento de Sistemas Energeticos, Facultad de Ingenieria, Universidad Nacional Autonoma de Mexico, Paseo Cuauhnahuac 8532. Jiutepec, Morelos (Mexico)

    2011-06-15

    Highlights: > Differences on reactivity with MCNPX and TRIPOLI-4 are negligible. > Fuel lattice and core criticality calculations were done. > A higher Doppler coefficient than coolant density coefficient. > Zirconium carbide is a better reflector than silicon carbide. > Adequate active height, radial size and reflector thickness were obtained. - Abstract: In this work the Monte Carlo codes MCNPX and TRIPOLI-4 were used to perform the criticality calculations of the fuel assembly and the core configuration of a gas-cooled fast reactor (GFR) concept, currently in development. The objective is to make contributions to the neutronic analysis of a gas-cooled fast reactor. In this study the fuel assembly is based on a hexagonal lattice of fuel-pins. The materials used are uranium and plutonium carbide as fuel, silicon carbide as cladding, and helium gas as coolant. Criticality calculations were done for a fuel assembly where the axial reflector thickness was varied in order to find the optimal thickness. In order to determine the best material to be used as a reflector, in the reactor core with neutrons of high energy spectrum, criticality calculations were done for three reflector materials: zirconium carbide, silicon carbide and natural uranium. It was found that the zirconium carbide provides the best neutron reflection. Criticality calculations using different active heights were done to determine the optimal height, and the reflector thickness was adjusted. Core criticality calculations were performed with different radius sizes to determine the active radial dimension of the core. A negative temperature coefficient of reactivity was verified for the fuel. The effect on reactivity produced by changes in the coolant density was also evaluated. We present the main neutronic characteristics of a preliminary fuel and core designs for the GFR concept. ENDF-VI cross-sections libraries were used in both the MCNPX and TRIPOLI-4 codes, and we verified that the obtained

  5. Depletion Analysis of Modular High Temperature Gas-cooled Reactor Loaded with LEU/Thorium Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Sonat Sen; Gilles Youinou

    2013-02-01

    Thorium based fuel has been considered as an option to uranium-based fuel, based on considerations of resource utilization (Thorium is more widely available when compared to Uranium). The fertile isotope of Thorium (Th-232) can be converted to fissile isotope U-233 by neutron capture during the operation of a suitable nuclear reactor such as High Temperature Gas-cooled Reactor (HTGR). However, the fertile Thorium needs a fissile supporter to start and maintain the conversion process such as U-235 or Pu-239. This report presents the results of a study that analyzed the thorium utilization in a prismatic HTGR, namely Modular High Temperature Gas-Cooled Reactor (MHTGR) that was designed by General Atomics (GA). The collected for the modeling of this design come from Chapter 4 of MHTGR Preliminary Safety Information Document that GA sent to Department of Energy (DOE) on 1995. Both full core and unit cell models were used to perform this analysis using SCALE 6.1 and Serpent 1.1.18. Because of the long mean free paths (and migration lengths) of neutrons in HTRs, using a unit cell to represent a whole core can be non-trivial. The sizes of these cells were set to match the spectral index between unit cell and full core domains. It was found that for the purposes of this study an adjusted unit cell model is adequate. Discharge isotopics and one-group cross-sections were delivered to the transmutation analysis team. This report provides documentation for these calculations

  6. Vortex Diode Analysis and Testing for Fluoride Salt-Cooled High-Temperature Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoder Jr, Graydon L [ORNL; Elkassabgi, Yousri M. [Texas A& M University, Kingsville; De Leon, Gerardo I. [Texas A& M University, Kingsville; Fetterly, Caitlin N. [Texas A& M University, Kingsville; Ramos, Jorge A. [Texas A& M University, Kingsville; Cunningham, Richard Burns [University of Tennessee, Knoxville (UTK)

    2012-02-01

    Fluidic diodes are presently being considered for use in several fluoride salt-cooled high-temperature reactor designs. A fluidic diode is a passive device that acts as a leaky check valve. These devices are installed in emergency heat removal systems that are designed to passively remove reactor decay heat using natural circulation. The direct reactor auxiliary cooling system (DRACS) uses DRACS salt-to-salt heat exchangers (DHXs) that operate in a path parallel to the core flow. Because of this geometry, under normal operating conditions some flow bypasses the core and flows through the DHX. A flow diode, operating in reverse direction, is-used to minimize this flow when the primary coolant pumps are in operation, while allowing forward flow through the DHX under natural circulation conditions. The DRACSs reject the core decay heat to the environment under loss-of-flow accident conditions and as such are a reactor safety feature. Fluidic diodes have not previously been used in an operating reactor system, and therefore their characteristics must be quantified to ensure successful operation. This report parametrically examines multiple design parameters of a vortex-type fluidic diode to determine the size of diode needed to reject a particular amount of decay heat. Additional calculations were performed to size a scaled diode that could be tested in the Oak Ridge National Laboratory Liquid Salt Flow Loop. These parametric studies have shown that a 152.4 mm diode could be used as a test article in that facility. A design for this diode is developed, and changes to the loop that will be necessary to test the diode are discussed. Initial testing of a scaled flow diode has been carried out in a water loop. The 150 mm diode design discussed above was modified to improve performance, and the final design tested was a 171.45 mm diameter vortex diode. The results of this testing indicate that diodicities of about 20 can be obtained for diodes of this size. Experimental

  7. Metal fuel development and verification for prototype generation- IV Sodium- Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chan Bock; Cheon, Jin Sik; Kim, Sung Ho; Park, Jeong Yong; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Metal fuel is being developed for the prototype generation-IV sodium-cooled fast reactor (PGSFR) to be built by 2028. U-Zr fuel is a driver for the initial core of the PGSFR, and U -transuranics (TRU)-Zr fuel will gradually replace U-Zr fuel through its qualification in the PGSFR. Based on the vast worldwide experiences of U-Zr fuel, work on U-Zr fuel is focused on fuel design, fabrication of fuel components, and fuel verification tests. U-TRU-Zr fuel uses TRU recovered through pyroelectrochemical processing of spent PWR (pressurized water reactor) fuels, which contains highly radioactive minor actinides and chemically active lanthanide or rare earth elements as carryover impurities. An advanced fuel slug casting system, which can prevent vaporization of volatile elements through a control of the atmospheric pressure of the casting chamber and also deal with chemically active lanthanide elements using protective coatings in the casting crucible, was developed. Fuel cladding of the ferritic-martensitic steel FC92, which has higher mechanical strength at a high temperature than conventional HT9 cladding, was developed and fabricated, and is being irradiated in the fast reactor.

  8. Update on Small Modular Reactors Dynamics System Modeling Tool -- Molten Salt Cooled Architecture

    Energy Technology Data Exchange (ETDEWEB)

    Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Cetiner, Sacit M. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Fugate, David L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Qualls, A L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Borum, Robert C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Chaleff, Ethan S. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Rogerson, Doug W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Batteh, John J. [Modelon Corporation (Sweden); Tiller, Michael M. [Xogeny Corporation, Canton, MI (United States)

    2014-08-01

    The Small Modular Reactor (SMR) Dynamic System Modeling Tool project is in the third year of development. The project is designed to support collaborative modeling and study of various advanced SMR (non-light water cooled) concepts, including the use of multiple coupled reactors at a single site. The objective of the project is to provide a common simulation environment and baseline modeling resources to facilitate rapid development of dynamic advanced reactor SMR models, ensure consistency among research products within the Instrumentation, Controls, and Human-Machine Interface (ICHMI) technical area, and leverage cross-cutting capabilities while minimizing duplication of effort. The combined simulation environment and suite of models are identified as the Modular Dynamic SIMulation (MoDSIM) tool. The critical elements of this effort include (1) defining a standardized, common simulation environment that can be applied throughout the program, (2) developing a library of baseline component modules that can be assembled into full plant models using existing geometry and thermal-hydraulic data, (3) defining modeling conventions for interconnecting component models, and (4) establishing user interfaces and support tools to facilitate simulation development (i.e., configuration and parameterization), execution, and results display and capture.

  9. Gas Cooled Fast Reactor Research and Development in the European Union

    Directory of Open Access Journals (Sweden)

    Richard Stainsby

    2009-01-01

    Full Text Available Gas-cooled fast reactor (GFR research is directed towards fulfilling the ambitious goals of Generation IV (Gen IV, that is, to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. The research is directed towards developing the GFR as an economic electricity generator, with good safety and sustainability characteristics. Fast reactors maximise the usefulness of uranium resources by breeding plutonium and can contribute to minimising both the quantity and radiotoxicity nuclear waste by actinide transmutation in a closed fuel cycle. Transmutation is particularly effective in the GFR core owing to its inherently hard neutron spectrum. Further, GFR is suitable for hydrogen production and process heat applications through its high core outlet temperature. As such GFR can inherit the non-electricity applications that will be developed for thermal high temperature reactors in a sustainable manner. The Euratom organisation provides a route by which researchers in all European states, and other non-European affiliates, can contribute to the Gen IV GFR system. This paper summarises the achievements of Euratom's research into the GFR system, starting with the 5th Framework programme (FP5 GCFR project in 2000, through FP6 (2005 to 2009 and looking ahead to the proposed activities within the 7th Framework Programme (FP7.

  10. Final report-passive safety optimization in liquid sodium-cooled reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Cahalana, J. E.; Hahn, D.; Nuclear Engineering Division; Korea Atomic Energy Research Inst.

    2007-08-13

    This report summarizes the results of a three-year collaboration between Argonne National Laboratory (ANL) and the Korea Atomic Energy Research Institute (KAERI) to identify and quantify the performance of innovative design features in metallic-fueled, sodium-cooled fast reactor designs. The objective of the work was to establish the reliability and safety margin enhancements provided by design innovations offering significant potential for construction, maintenance, and operating cost reductions. The project goal was accomplished with a combination of advanced model development (Task 1), analysis of innovative design and safety features (Tasks 2 and 3), and planning of key safety experiments (Task 4). Task 1--Computational Methods for Analysis of Passive Safety Design Features: An advanced three-dimensional subassembly thermal-hydraulic model was developed jointly and implemented in ANL and KAERI computer codes. The objective of the model development effort was to provide a high-accuracy capability to predict fuel, cladding, coolant, and structural temperatures in reactor fuel subassemblies, and thereby reduce the uncertainties associated with lower fidelity models previously used for safety and design analysis. The project included model formulation, implementation, and verification by application to available reactor tests performed at EBR-II. Task 2--Comparative Analysis and Evaluation of Innovative Design Features: Integrated safety assessments of innovative liquid metal reactor designs were performed to quantify the performance of inherent safety features. The objective of the analysis effort was to identify the potential safety margin enhancements possible in a sodium-cooled, metal-fueled reactor design by use of passive safety mechanisms to mitigate low-probability accident consequences. The project included baseline analyses using state-of-the-art computational models and advanced analyses using the new model developed in Task 1. Task 3--Safety

  11. Air Corrosivity in U.S. Outdoor-Air-Cooled Data Centers is Similar to That in Conventional Data Centers

    Energy Technology Data Exchange (ETDEWEB)

    Coles, Henry C.; Han, Taewon; Price, Phillip N.; Gadgil, Ashok J.; Tschudi, William F.

    2011-07-17

    There is a concern that environmental-contamination caused corrosion may negatively affect Information Technology (IT) equipment reliability. Nineteen data centers in the United States and two in India were evaluated using Corrosion Classification Coupons (CCC) to assess environmental air quality as it may relate IT equipment reliability. The data centers were of two basic types: closed and outside-air cooled. A closed data center provides cool air to the IT equipment using air conditioning in which only a small percent age of the recirculation air is make-up air continuously supplied from outside to meet human health requirements. An outside-air cooled data center uses outside air directly as the primary source for IT equipment cooling. Corrosion measuring coupons containing copper and silver metal strips were placed in both closed and outside-air cooled data centers. The coupons were placed at each data center (closed and outside-air cooled types) with the location categorized into three groups: (1) Outside - coupons sheltered, located near or at the supply air inlet, but located before any filtering, (2) Supply - starting just after initial air filtering continuing inside the plenums and ducts feeding the data center rooms, and (3) Inside located inside the data center rooms near the IT equipment. Each coupon was exposed for thirty days and then sent to a laboratory for a corrosion rate measurement analysis. The goal of this research was to investigate whether gaseous contamination is a concern for U.S. data center operators as it relates to the reliability of IT equipment. More specifically, should there be an increased concern if outside air for IT equipment cooling is used To begin to answer this question limited exploratory measurements of corrosion rates in operating data centers in various locations were undertaken. This study sought to answer the following questions: (1) What is the precision of the measurements (2) What are the approximate statistical

  12. Development of a water-mist cooling system: A 12,500 Kcal/h air-cooled chiller

    Directory of Open Access Journals (Sweden)

    Chung-Neng Huang

    2015-11-01

    Full Text Available Global warming and energy exhaustion problems are becoming a severe problems, of which energy conservation and carbon reduction are the most critical. Between 40% and 48% of the total electricity used in a building is consumed by air conditioning systems. The development of a supersonic water-misting cooling system with a fuzzy control system is proposed to optimize existing condenser noise, space, and energy consumption, as well as to address problems with cooling capacity resulting from improper control between compressors and condensers. An experimental platform was established for conducting tests, observing cooling efficiencies, and calculating power saving statuses. Comparing the observed cooling efficiency, a temperature difference of 5.4 °C was determined before and after the application; this is significant regarding efficiency. The method produces no pollution or water accumulation. When compared with fixed frequency air-cooled water chillers, an exceptional energy saving of 25% was observed. The newly developed supersonic mist-cooled chiller is an excellent solution to increasing water and electricity fees.

  13. Toward a Mechanistic Source Term in Advanced Reactors: Characterization of Radionuclide Transport and Retention in a Sodium Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Brunett, Acacia J.; Bucknor, Matthew; Grabaskas, David

    2016-04-17

    A vital component of the U.S. reactor licensing process is an integrated safety analysis in which a source term representing the release of radionuclides during normal operation and accident sequences is analyzed. Historically, source term analyses have utilized bounding, deterministic assumptions regarding radionuclide release. However, advancements in technical capabilities and the knowledge state have enabled the development of more realistic and best-estimate retention and release models such that a mechanistic source term assessment can be expected to be a required component of future licensing of advanced reactors. Recently, as part of a Regulatory Technology Development Plan effort for sodium cooled fast reactors (SFRs), Argonne National Laboratory has investigated the current state of knowledge of potential source terms in an SFR via an extensive review of previous domestic experiments, accidents, and operation. As part of this work, the significant sources and transport processes of radionuclides in an SFR have been identified and characterized. This effort examines all stages of release and source term evolution, beginning with release from the fuel pin and ending with retention in containment. Radionuclide sources considered in this effort include releases originating both in-vessel (e.g. in-core fuel, primary sodium, cover gas cleanup system, etc.) and ex-vessel (e.g. spent fuel storage, handling, and movement). Releases resulting from a primary sodium fire are also considered as a potential source. For each release group, dominant transport phenomena are identified and qualitatively discussed. The key product of this effort was the development of concise, inclusive diagrams that illustrate the release and retention mechanisms at a high level, where unique schematics have been developed for in-vessel, ex-vessel and sodium fire releases. This review effort has also found that despite the substantial range of phenomena affecting radionuclide release, the

  14. A Cylindrical Shielding Design Concept for the Prototype Gen-IV Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yun, Sunghwan; Kim, Sang Ji [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    In the Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR), a metal fueled, blanket-free, pool type SFR concept is adopted to acquire the inherent safety characteristics and high proliferation-resistance. In the pool type fast reactor, the intermediate heat exchangers (IHXs), which transfer heat from the primary sodium pool to a secondary sodium loop, are placed inside of the reactor vessel. Hence, secondary sodium passing the IHXs can be radioactivated by a {sup 23}Na(n,g){sup 24}Na reaction, and radioactivated secondary sodium causes a significant dose in the Steam Generator Building (SGB). Therefore, a typical core of a pool type fast reactor is usually surrounded by a massive quantity of shields. In addition, the blanket composed of depleted uranium plays a role as superior shielding material; a significant increase in shields is required in the blanket-free pool type SFR. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR. In a conventional shielding design, massive axial shields are required to prevent irradiation of secondary sodium passing IHXs and they should be replaced according to the subassembly replacement in spite of negligible depletion of the shielding material. The proposed shielding design concept minimizes the quantity of shields without their replacement. In this paper, a new cylindrical shielding design concept is proposed for a blanket-free pool type SFR such as a PGSFR. The proposed design concept satisfied the dose limit in the steam generator building successfully without introducing a large quantity of B{sub 4}C shielding inside the subassembly.

  15. Thermal Aspects of Using Alternative Nuclear Fuels in Supercritical Water-Cooled Reactors

    Science.gov (United States)

    Grande, Lisa Christine

    A SuperCritical Water-cooled Nuclear Reactor (SCWR) is a Generation IV concept currently being developed worldwide. Unique to this reactor type is the use of light-water coolant above its critical point. The current research presents a thermal-hydraulic analysis of a single fuel channel within a Pressure Tube (PT)-type SCWR with a single-reheat cycle. Since this reactor is in its early design phase many fuel-channel components are being investigated in various combinations. Analysis inputs are: steam cycle, Axial Heat Flux Profile (AHFP), fuel-bundle geometry, and thermophysical properties of reactor coolant, fuel sheath and fuel. Uniform and non-uniform AHFPs for average channel power were applied to a variety of alternative fuels (mixed oxide, thorium dioxide, uranium dicarbide, uranium nitride and uranium carbide) enclosed in an Inconel-600 43-element bundle. The results depict bulk-fluid, outer-sheath and fuel-centreline temperature profiles together with the Heat Transfer Coefficient (HTC) profiles along the heated length of fuel channel. The objective is to identify the best options in terms of fuel, sheath material and AHFPS in which the outer-sheath and fuel-centreline temperatures will be below the accepted temperature limits of 850°C and 1850°C respectively. The 43-element Inconel-600 fuel bundle is suitable for SCWR use as the sheath-temperature design limit of 850°C was maintained for all analyzed cases at average channel power. Thoria, UC2, UN and UC fuels for all AHFPs are acceptable since the maximum fuel-centreline temperature does not exceed the industry accepted limit of 1850°C. Conversely, the fuel-centreline temperature limit was exceeded for MOX at all AHFPs, and UO2 for both cosine and downstream-skewed cosine AHFPs. Therefore, fuel-bundle modifications are required for UO2 and MOX to be feasible nuclear fuels for SCWRs.

  16. Design and Test of Advanced Thermal Simulators for an Alkali Metal-Cooled Reactor Simulator

    Science.gov (United States)

    Garber, Anne E.; Dickens, Ricky E.

    2011-01-01

    The Early Flight Fission Test Facility (EFF-TF) at NASA Marshall Space Flight Center (MSFC) has as one of its primary missions the development and testing of fission reactor simulators for space applications. A key component in these simulated reactors is the thermal simulator, designed to closely mimic the form and function of a nuclear fuel pin using electric heating. Continuing effort has been made to design simple, robust, inexpensive thermal simulators that closely match the steady-state and transient performance of a nuclear fuel pin. A series of these simulators have been designed, developed, fabricated and tested individually and in a number of simulated reactor systems at the EFF-TF. The purpose of the thermal simulators developed under the Fission Surface Power (FSP) task is to ensure that non-nuclear testing can be performed at sufficiently high fidelity to allow a cost-effective qualification and acceptance strategy to be used. Prototype thermal simulator design is founded on the baseline Fission Surface Power reactor design. Recent efforts have been focused on the design, fabrication and test of a prototype thermal simulator appropriate for use in the Technology Demonstration Unit (TDU). While designing the thermal simulators described in this paper, effort were made to improve the axial power profile matching of the thermal simulators. Simultaneously, a search was conducted for graphite materials with higher resistivities than had been employed in the past. The combination of these two efforts resulted in the creation of thermal simulators with power capacities of 2300-3300 W per unit. Six of these elements were installed in a simulated core and tested in the alkali metal-cooled Fission Surface Power Primary Test Circuit (FSP-PTC) at a variety of liquid metal flow rates and temperatures. This paper documents the design of the thermal simulators, test program, and test results.

  17. Study of evaluation methods for in-vessel corium retention through external vessel cooling and safety of reactor cavity

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hoon; Chang, Soon Heung; Kim, Soo Hyung; Kim, Kee Poong; Lee, Hyoung Wook; Jang, Kwang Keol; Jeong, Yong Hoon; Kim, Sang Jin; Lee, Seong Jin [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Park, Jae Hong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    2001-03-15

    In this work, assessment system for methodology for reactor pressure vessel integrity is developed. Assessment system is make up of severe accident assessment code which can calculate the conditions of plant and structural analysis code which can assess the integrity of reactor vessel using given plant conditions. An assessment of cavity flooding using containment spray system has been done. As a result, by the containment spray, cavity can be flooded successfully and CCI can be reduced. The technical backgrounds for external vessel cooling and corium cooling on the cavity are summarized and provided in this report.

  18. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    Science.gov (United States)

    Afifah, Maryam; Miura, Ryosuke; Su'ud, Zaki; Takaki, Naoyuki; Sekimoto, H.

    2015-09-01

    Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don't need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

  19. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

    This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.

  20. Influence of Different Vortex Generators on Heat Transfer in Direct Air-Cooled Condensers

    Institute of Scientific and Technical Information of China (English)

    ZHOU Guobing; YANG Laishun

    2012-01-01

    With the capacity of air cooling turbines increasing, the air-cooled steam condenser has been developed to the single fiat tube. Fig.1 shows the studied single-row wavy-finned flat tube. Vortex generators are more and more used in oil and chemical industry, power generation, and refrigeration industry in the heat exchanger. The analyses are performed to investigate the effect of vortex generators on pressure drop and heat transfer of single-row wavy-finned fiat tubes of direct air-cooled condensers.

  1. Discussion on Feasibility and Economy of 1000-MW Ultra-Supercritical Air-Cooling Unit

    Institute of Scientific and Technical Information of China (English)

    Zhu Jun; Wang Yunze; Jin Wen

    2007-01-01

    @@ Based on analysis on types and features of domestically made 1000-MW large ultra supercritical steam turbine and 600-MW air cooling steam turbine,the author puts forward that 1000-MW ultra supercritical air-cooling turbine can be assembled with high and medium pressure cylinder modules of 1000-MW ultra-supercritical steam turbine and low-pressure cylinder module of 600-MW tow-cylinder and tow-exhaust air-cooling turbine.In addition,the economy of the assembled turbine is discussed, and designing considerations and issues need to be furtherstudied are proposed as well.

  2. Development of gas cooled reactors and experimental setup of high temperature helium loop for in-pile operation

    Energy Technology Data Exchange (ETDEWEB)

    Miletić, Marija, E-mail: marija_miletic@live.com [Czech Technical University in Prague, Prague (Czech Republic); Fukač, Rostislav, E-mail: fuk@cvrez.cz [Research Centre Rez Ltd., Rez (Czech Republic); Pioro, Igor, E-mail: Igor.Pioro@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada); Dragunov, Alexey, E-mail: Alexey.Dragunov@uoit.ca [University of Ontario Institute of Technology, Oshawa (Canada)

    2014-09-15

    Highlights: • Gas as a coolant in Gen-IV reactors, history and development. • Main physical parameters comparison of gas coolants: carbon dioxide, helium, hydrogen with water. • Forced convection in turbulent pipe flow. • Gas cooled fast reactor concept comparisons to very high temperature reactor concept. • High temperature helium loop: concept, development, mechanism, design and constraints. - Abstract: Rapidly increasing energy and electricity demands, global concerns over the climate changes and strong dependence on foreign fossil fuel supplies are powerfully influencing greater use of nuclear power. In order to establish the viability of next-generation reactor concepts to meet tomorrow's needs for clean and reliable energy production the fundamental research and development issues need to be addressed for the Generation-IV nuclear-energy systems. Generation-IV reactor concepts are being developed to use more advanced materials, coolants and higher burn-ups fuels, while keeping a nuclear reactor safe and reliable. One of the six Generation-IV concepts is a very high temperature reactor (VHTR). The VHTR concept uses a graphite-moderated core with a once-through uranium fuel cycle, using high temperature helium as the coolant. Because helium is naturally inert and single-phase, the helium-cooled reactor can operate at much higher temperatures, leading to higher efficiency. Current VHTR concepts will use fuels such as uranium dioxide, uranium carbide, or uranium oxycarbide. Since some of these fuels are new in nuclear industry and due to their unknown properties and behavior within VHTR conditions it is very important to address these issues by investigate their characteristics within conditions close to those in VHTRs. This research can be performed in a research reactor with in-pile helium loop designed and constructed in Research Center Rez Ltd. One of the topics analyzed in this article are also physical characteristic and benefits of gas

  3. A Management Strategy for the Heavy Water Reflector Cooling System of HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, H. S.; Park, Y. C.; Lim, S. P. (and others)

    2007-11-15

    Heavy water is used as the reflector and the moderator of the HANARO research reactor. After over 10 years operation since first criticality in 1995 there arose some operational issues related with the tritium. A task force team(TFT) has been operated for 1 year since September 2006 to study and deduce resolutions of the issues concerning the tritium and the degradation of heavy water in the HANARO reflector system. The TFT drew many recommendations on the hardware upgrade, tritium containing air control, heavy water quality management, waste management, and tritium measurement system upgrade.

  4. TRACG-CFD analysis of ESBWR reactor water cleanup shutdown cooling system mixing coefficient

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo, J. [UNAM, Facultad de Ingenieria, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Marquino, W.; Mistreanu, A.; Yang, J., E-mail: euqrop@hotmail.com [General Electric Hitachi Nuclear Energy, Wilmington, 28401 North Carolina (United States)

    2015-09-15

    The ESBWR is a 1520 nominal [M We] Generation III+ natural circulation boiling water reactor designed to high levels of safety utilizing features that have been successfully used before in operating BWRs, as well as standard features common to A BWR. In September of 2014, the US NRC has certified the ESBWR design for use in the USA. The RWCU/Sdc is an auxiliary system for the ESBWR nuclear island. Basic functions it performs include purifying the reactor coolant during normal operation and shutdown and providing shutdown cooling and cooldown to cold shutdown conditions. The performance of the RWCU system during shutdown cooling is directly related to the temperature of the water removed through the outlets, which is coupled with the vessel and F W temperatures through a thermal mixing coefficient. The complex three-dimensional (3-D) geometry of the BWR downcomer and lower plenum has a great impact on the flow mixing. Only a fine mesh technique like CFD can predict the 3-D temperature distribution in the RPV during shutdown and provide the RWCU/Sdc system inlet temperature. Plant shutdown is an unsteady event by nature and was modeled as a succession of CFD steady-state simulations. It is required to establish the mixing coefficient (which is a function of the heat balance and the core flow) during the operation of the RWCU system in the multiple shutdown cooling modes, and therefore a range of core flows needs to be estimated using quasi steady states obtained with TRACG. The lower end of that range is obtained from a system with minimal power decay heat and core flow; while the higher end corresponds to the power at the beginning of RWCU/Sdc operation when the cooldown is transferred to the RWCU/Sdc after the initial depressurization via the turbine bypass valves. Because the ESBWR RWCU/Sdc return and suction designs provide good mixing, the uniform mixing energy balance was found to be an adequate alternative for deriving the mixing coefficient. The CFD mass flow

  5. Use of Air2Air Technology to Recover Fresh-Water from the Normal Evaporative Cooling Loss at Coal-Based Thermoelectric Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Ken Mortensen

    2009-06-30

    This program was undertaken to build and operate the first Air2Air{trademark} Water Conservation Cooling Tower at a power plant, giving a validated basis and capability for water conservation by this method. Air2Air{trademark} water conservation technology recovers a portion of the traditional cooling tower evaporate. The Condensing Module provides an air-to-air heat exchanger above the wet fill media, extracting the heat from the hot saturated moist air leaving in the cooling tower and condensing water. The rate of evaporate water recovery is typically 10%-25% annually, depending on the cooling tower location (climate).

  6. Safety and licensing of MHTGR (Modular High Temperature Gas Cooled Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Silady, F.A.; Millunzi, A.C.; Kelley, A.P. Jr.; Cunliffe, J.

    1987-07-01

    The Modular High Temperature Gas Cooled Reactor (MHTGR) design meets stringent top-level regulatory and user safety requirements that require that the normal and off-normal operation of the plant not disturb the public's day-to-day activities. Quantitative, top-level regulatory criteria have been specified from US NRC and EPA sources to guide the design. The user/utility group has further specified that these criteria be met at the plant boundary. The focus of the safety approach has then been centered on retaining the radionuclide inventory within the fuel by removing core heat, controlling chemical attack, and by controlling heat generation. The MHTGR is shown to passively meet the stringent requirements with margin. No operator action is required and the plant is insensitive to operator error.

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

    Science.gov (United States)

    Aures, Alexander; Bostelmann, Friederike; Zwermann, Winfried; Velkov, Kiril

    2016-03-01

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

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

    Directory of Open Access Journals (Sweden)

    Aures Alexander

    2016-01-01

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

  9. Development of the Sodium-cooled Fast Reactor R and D and Technology Monitoring System

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dong Uk; Won, Byung Chool; Kim, Young In; Hahn, Do Hee

    2008-01-15

    This study presents a R and D performance monitoring system that is applicable for managing the generation IV sodium-cooled fast reactor development. The prime goal of this system is to furnish project manager with reliable and accurate information of status of progress, performance and resource allocation, and attain traceability and visibility of project implementation for effective project management. In this study, the work breakdown structure, the related schedule and the expected outputs were established to derive the interfaces between projects and the above parameters was loaded PCs. The R and D performance monitoring system is composed of about 750 R and D activities within 'Development of Basic Key Technologies for Gen IV SFR' project in 2007. The Microsoft Project Professional software was used to monitor the progress, evaluate the results and analyze the resource distribution to activities.

  10. Distribution and abundance of submerged aquatic macrophytes in a reactor cooling reservoir

    Energy Technology Data Exchange (ETDEWEB)

    Grace, J.B.

    1977-08-01

    Measurements of ash-free dry weight were used to characterize the effects of a heated effluent on submerged macrophytes in a reactor cooling reservoir. The species which were most abundant during the summers of 1974 and 1975 were Myriophyllum spicatum L. and Eleocharis acicularis (L.) R. and S. Examination of the vertical distribution of the shoot biomass of Myriophyllum revealed that plants in heated areas grew closer to the water surface than plants in unheated areas. The biomass of the second most abundant species, Eleocharis acicularis, was less at 0.5 m depths in heated areas (more than 5C/sup 0/ warmer than unheated areas) than at equal depths in unheated areas. Species diversity was greater at heated locations because of a greater equitability (i.e., evenness of distribution of biomass) among species.

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

    Science.gov (United States)

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

    1973-01-01

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

  12. Simultaneous approach for simulation of a high-temperature gas-cooled reactor

    Institute of Scientific and Technical Information of China (English)

    Yang CHEN; Jiang-hong YOU; Zhi-jiang SHAO; Ke-xin WANG; Ji-xin QIAN

    2011-01-01

    The simulation of a high-temperature gas-cooled reactor pebble-bed module (HTR-PM) plant is discussed.This lumped parameter model has the form of a set differential algebraic equations (DAEs) that include stiff equations to model point neutron kinetics.The nested approach is the most common method to solve DAE,but this approach is very expensive and time-consuming due to inner iterations.This paper deals with an alternative approach in which a simultaneous solution method is used.The DAEs are discretized over a time horizon using collocation on finite elements,and Radau collocation points are applied.The resulting nonlinear algebraic equations can be solved by existing solvers.The discrete algorithm is discussed in detail; both accuracy and stability issues are considered.Finally,the simulation results are presented to validate the efficiency and accuracy of the simultaneous approach that takes much less time than the nested one.

  13. High Temperature Gas-Cooled Reactor Projected Markets and Preliminary Economics

    Energy Technology Data Exchange (ETDEWEB)

    Larry Demick

    2011-08-01

    This paper summarizes the potential market for process heat produced by a high temperature gas-cooled reactor (HTGR), the environmental benefits reduced CO2 emissions will have on these markets, and the typical economics of projects using these applications. It gives examples of HTGR technological applications to industrial processes in the typical co-generation supply of process heat and electricity, the conversion of coal to transportation fuels and chemical process feedstock, and the production of ammonia as a feedstock for the production of ammonia derivatives, including fertilizer. It also demonstrates how uncertainties in capital costs and financial factors affect the economics of HTGR technology by analyzing the use of HTGR technology in the application of HTGR and high temperature steam electrolysis processes to produce hydrogen.

  14. Molecular Dynamics Simulations of Aqueous and Confined Systems Relevant to the Supercritical Water Cooled Nuclear Reactor

    Science.gov (United States)

    Kallikragas, Dimitrios Theofanis

    Supercritical water (SCW) is the intended heat transfer fluid and potential neutron moderator in the proposed GEN-IV Supercritical Water Cooled Reactor (SCWR). The oxidative environment poses challenges in choosing appropriate design materials, and the behaviour of SCW within crevices of the passivation layer is needed for developing a corrosion control strategy to minimize corrosion. Molecular Dynamics simulations have been employed to obtain diffusion coefficients, coordination number and surface density characteristics, of water and chloride in nanometer-spaced iron hydroxide surfaces. Diffusion models for hydrazine are evaluated along with hydration data. Results demonstrate that water is more likely to accumulate on the surface at low density conditions. The effect of confinement on the water structure diminishes as the gap size increases. The diffusion coefficient of chloride decreases with larger surface spacing. Clustering of water at the surface implies that the SCWR will be most susceptible to pitting corrosion and stress corrosion cracking.

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

    Science.gov (United States)

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

    2011-03-01

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

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

    Science.gov (United States)

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

    2013-09-03

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

  17. Temperature control of bench-scaled batch reactor equipped with a monofluid heating/cooling system

    Science.gov (United States)

    Teng, Hai-peng; Song, Yi-ming

    2014-04-01

    An advanced control concept, Predictive Functional Control (PFC), is applied for temperature control of a bench-scaled batch reactor equipped with monofluid heating/cooling system. First principles process models are developed. Based on achieved models, significant process variables, which are difficult or impossible to measure online, are estimated from easily measured variables, and cascade PFC control strategy has been projected and implemented in Matlab R14. The dynamics of individual subunits is explicitly taken into consideration by internal model in the control algorithms, and model uncertainty, various process disturbances are compensated by modification of internal model. The experimental results present an excellent capability of tracking the set point, and the success of PFC technique as a process control paradigm is illustratively demonstrated.

  18. Use of Distribution Devices for Hydraulic Profiling of Coolant Flow in Core Gas-cooled Reactors

    Directory of Open Access Journals (Sweden)

    A. A. Satin

    2014-01-01

    Full Text Available In setting up a reactor plant for the transportation-power module of the megawatt class an important task is to optimize the path of flow, i.e. providing moderate hydraulic resistance, uniform distribution of the coolant. Significant contribution to the hydraulic losses makes one selected design of the coolant supplies. It is, in particular, hemispherical or semi-elliptical shape of the supply reservoir, which is selected to reduce its mass, resulting in the formation of torusshaped vortex in the inlet manifold, that leads to uneven coolant velocity at the inlet into the core, the flow pulsations, hydraulic losses.To control the flow redistribution in the core according to the level of energy are used the switchgear - deflectors installed in a hemispherical reservoir supplying coolant to the fuel elements (FE of the core of gas-cooled reactor. This design solution has an effect on the structure of the flow, rate in the cooling duct, and the flow resistance of the collector.In this paper we present the results of experiments carried out on the gas dynamic model of coolant paths, deflectors, and core, comprising 55 fuel rod simulators. Numerical simulation of flow in two-parameter model, using the k-ε turbulence model, and the software package ANSYS CFX v14.0 is performed. The paper demonstrates that experimental results are in compliance with calculated ones.The results obtained suggest that the use of switchgear ensures a coolant flow balance directly at the core inlet, thereby providing temperature reduction of fuel rods with a uniform power release in the cross-section. Considered options to find constructive solutions for deflectors give an idea to solve the problem of reducing hydraulic losses in the coolant paths, to decrease pulsation components of flow in the core and length of initial section of flow stabilization.

  19. Dynamic Free Cooling. Efficient and energy saving air conditioner for datahotels; Dynamic Free Cooling. Efficient en energeibesparend airconditioningsysteem voor datacenters

    Energy Technology Data Exchange (ETDEWEB)

    Havenaar, D.

    2009-01-15

    Servers in data centres produce a large amount of heat, which has to be removed by ventilation and air conditioning systems. These systems have a vast energy consumption. Increasing energy costs and limited resources of available electricity are forcing the data centre industry to the use of energy efficient technical equipment. Dynamic Free Cooling is a contrN concept for data centre's air conditioning systems combining Hybrid Indirect Free Cooling Precision Air Conditiening Units, Fan Speed Controlled Dry Coolers and Speed Controlled Central Pumps to highly efficient precision cooling system. All system components are centrally controlled to minimize overall energy consumption depending on the Ambient Temperature and the Room Load Status. [Dutch] Servers in datacenters produceren een grote hoeveelheid warmte die moet worden afgevoerd door middel van ventilatie- en airconditioningsystemen. Deze systemen gebruiken een forse hoeveelheid elektrische energie. Toenemende energiekosten en de beperkte beschikbaarheid van elektriciteit zorgen ervoor dat datacenters steeds meer overgaan op het installeren van energie-efficiente systemen. Dynamic Free Cooling is een regelconcept voor airconditioningsystemen in combinatie met hybride indirecte vrije koeling in precisie-airconditioningunits, ventilatorgeregetde droge koelers en toerengeregetde centrale pompsystemen. AI deze systeemcomponenten worden centraal geregeld om het totale energiegebruik, afhankelijk van de omgevingstemperatuur en de koellast te minimaliseren.

  20. Measurement of Turbulent Flow Phenomena for the Lower Plenum of a Prismatic Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hugh M. McIlroy Jr.; Donald M. McEligot; Robert J. Pink; Keith G. Condie; Glenn E. McCreery

    2007-09-01

    Mean velocity field and turbulence data are presented for flow phenomena in a lower plenum of a typical prismatic gas-cooled reactor (GCR), such as in a Very High Temperature Reactor (VHTR) concept. In preparation for design, safety analyses and licensing, research has begun on readying the computational tools that will be needed to predict the thermal-hydraulics behavior of the reactor design. Fluid dynamics experiments have been designed and built to develop benchmark databases for the assessment of computational fluid dynamics (CFD) codes and their turbulence models for a typical VHTR plenum geometry in the limiting case of negligible buoyancy and constant fluid properties. This experiment has been proposed as a “Standard Problem” for assessing advanced reactor (CFD) analysis tools. Present results concentrate on the region of the plenum near its far reflector wall (away from the outlet duct). The flow in the lower plenum can locally be considered as multiple jets into a confined cross flow - with obstructions. A model of the lower plenum has been fabricated and scaled to the geometric dimensions of the Next Generation Nuclear Plant (NGNP) Point Design. The model consists of a row of full circular posts along its centerline with half-posts on the two parallel walls to induce flow features somewhat comparable to those expected from the staggered parallel rows of posts in the reactor design. Posts, side walls and end walls are fabricated from clear, fused quartz to match the refractive-index of the working fluid so that optical techniques may be employed for the measurements. The experiments were conducted in the Matched-Index-of-Refraction (MIR) Facility at the Idaho National Laboratory (INL). The benefit of the MIR technique is that it permits optical measurements to determine complex flow characteristics in passages and around objects to be obtained without locating a disturbing transducer in the flow field and without distortion of the optical paths. The

  1. Validation of SCALE and the TRITON Depletion Sequence for Gas-Cooled Reactor Analysis

    Energy Technology Data Exchange (ETDEWEB)

    DeHart, Mark D [ORNL; Pritchard, Megan L [ORNL

    2008-01-01

    The very-high-temperature reactor (VHTR) is an advanced reactor concept that uses graphite-moderated fuel and helium gas as a coolant. At present there are two primary VHTR reactor designs under consideration for development: in the pebble-bed reactor, a core is loaded with 'pebbles' consisting of 6 cm diameter spheres, while in a high-temperature gas-cooled reactor, fuel rods are placed within prismatic graphite blocks. In both systems, fuel elements (spheres or rods) are comprised of tristructural-isotropic (TRISO) fuel particles. The TRISO particles are either dispersed in the matrix of a graphite pebble for the pebble-bed design or molded into compacts/rods that are then inserted into the hexagonal graphite blocks for the prismatic concept. Two levels of heterogeneity exist in such fuel designs: (1) microspheres of TRISO particles dispersed in a graphite matrix of a cylindrical or spherical shape, and (2) neutron interactions at the rod-to-rod or sphere-to-sphere level. Such double heterogeneity (DH) provides a challenge to multigroup cross-section processing methods, which must treat each level of heterogeneity separately. A new capability to model doubly heterogeneous systems was added to the SCALE system in the release of Version 5.1. It was included in the control sequences CSAS and CSAS6, which use the Monte Carlo codes KENO V.a and KENO-VI, respectively, for three-dimensional neutron transport analyses and in the TRITON sequence, which uses the two-dimensional lattice physics code NEWT along with both versions of KENO for transport and depletion analyses. However, the SCALE 5.1 version of TRITON did not support the use of the DH approach for depletion. This deficiency has been addressed, and DH depletion will be available as an option in the upcoming release of SCALE 6. At present Oak Ridge National Laboratory (ORNL) staff are developing a set of calculations that may be used to validate SCALE for DH calculations. This paper discusses the

  2. Cooling Rates of Humans in Air and in Water: An Experiment

    Science.gov (United States)

    Bohren, Craig F.

    2012-12-01

    In a previous article I analyzed in detail the physical factors resulting in greater cooling rates of objects in still water than in still air, emphasizing cooling of the human body. By cooling rate I mean the rate of decrease of core temperature uncompensated by metabolism. I concluded that the "correct ratio for humans is closer to 2 than to 10." To support this assertion I subsequently did experiments, which I report following a digression on hypothermia.

  3. Craft-joule project: air-cooled water LiBr absorption cooling machine of low capacity for air conditioning (ACABMA)

    Energy Technology Data Exchange (ETDEWEB)

    Oliva, A; Castro, J; Perez Segarra, C.D [Universitat Politecnica de Catalunya, Barcelona (Spain); Lucena, M.A [Instituto Nacional de Tecnica Aeroespecial (Spain)] (and others)

    2000-07-01

    The ACABMA (Air-Cooled water-LiBr Absorption cooling Machine of low capacity for Air- conditioning) project is a Craft-Joule Project within the framework of the Non Nuclear Energy Programme Joule III coordinated by the Centre Technologic de Transferencia de Calor (CTTC). The basic objective of this project is the development of a new air-cooled absorption cooling machine for air-conditioning, in the low power sector market. Making use of water-LiBr technology together with the air-cooling feature, it is possible to reach a better relationship between quality (in terms of performance, ecology, etc.) and price of such absorption machines, than the ones existing on the market. Air-cooling instead of water cooling saves installation costs specially in small systems and removes the demand for cooling water (an important aspect in Southern-European countries), thus increasing the possible application range. The main interest for the SME proposers is to take advantage of the increasing cooling demand in Europe, specially in southern countries. Another point of interest for the SME proposers is the development of a cheaper cooling and heating system in terms of energy and installation costs. In this moment the solar cooling systems are approx. 30% more expensive than the conventional ones. A cheaper absorption machine due to the air-cooling feature together with the possibility of energy savings due to low generator temperatures, that allow the absorption machine for solar applications or waste heat, will lead to solar cooling and heating systems more competitive to the conventional ones. In order to achieve the above mentioned goal, the following step are necessary and will be carried out in this project: i)solution of the air-cooling of the water-LiBr machine, the main problem that up to now has not allowed commercialization, ii)reduction of the size of the air-cooled elements of the machine in order to reduce the machine costs, iii)development of an efficient control

  4. Interim status report on lead-cooled fast reactor (LFR) research and development.

    Energy Technology Data Exchange (ETDEWEB)

    Tzanos, C. P.; Sienicki, J. J.; Moisseytsev, A.; Smith, C. F.; de Caro, M.; Halsey, W. G.; Li, N.; Hosemann, P.; Zhang, J.; Bolind, A.; LLNL; LANL; Univ. of Illinois

    2008-03-31

    This report discusses the status of Lead-Cooled Fast Reactor (LFR) research and development carried out during the first half of FY 2008 under the U.S. Department of Energy Generation IV Nuclear Energy Systems Initiative. Lead-Cooled Fast Reactor research and development has recently been transferred from Generation IV to the Reactor Campaign of the Global Nuclear Energy Partnership (GNEP). Another status report shall be issued at the end of FY 2008 covering all of the LFR activities carried out in FY 2008 for both Generation IV and GNEP. The focus of research and development in FY 2008 is an initial investigation of a concept for a LFR Advanced Recycling Reactor (ARR) Technology Pilot Plant (TPP)/demonstration test reactor (demo) incorporating features and operating conditions of the European Lead-cooled SYstem (ELSY) {approx} 600 MWe lead (Pb)-cooled LFR preconceptual design for the transmutation of waste and central station power generation, and which would enable irradiation testing of advanced fuels and structural materials. Initial scoping core concept development analyses have been carried out for a 100 MWt core composed of sixteen open-lattice 20 by 20 fuel assemblies largely similar to those of the ELSY preconceptual fuel assembly design incorporating fuel pins with mixed oxide (MOX) fuel, central control rods in each fuel assembly, and cooled with Pb coolant. For a cycle length of three years, the core is calculated to have a conversion ratio of 0.79, an average discharge burnup of 108 MWd/kg of heavy metal, and a burnup reactivity swing of about 13 dollars. With a control rod in each fuel assembly, the reactivity worth of an individual rod would need to be significantly greater than one dollar which is undesirable for postulated rod withdrawal reactivity insertion events. A peak neutron fast flux of 2.0 x 10{sup 15} (n/cm{sup 2}-s) is calculated. For comparison, the 400 MWt Fast Flux Test Facility (FFTF) achieved a peak neutron fast flux of 7.2 x 10{sup

  5. Army Gas-Cooled Reactor Systems Program. Operation of ML-1 reactor skid in GCRE: safety evaluation report

    Energy Technology Data Exchange (ETDEWEB)

    None

    1964-10-01

    The operation of the ML-1 reactor skid in the modified GCRE facility, utilizing the GCRE reactor coolant circulating and heat removal systems, is described. An evaluation of the safety considerations associated with this mode of operation indicates that the consequences of the maximum credible accident are less severe than those previously approved for operation of the ML-1 reactor at the ML-1 test site or for operation of the GCRE-I reactor in the GCRE facility.

  6. 16 CFR Appendix H to Part 305 - Cooling Performance and Cost for Central Air Conditioners

    Science.gov (United States)

    2010-01-01

    ... 16 Commercial Practices 1 2010-01-01 2010-01-01 false Cooling Performance and Cost for Central Air Conditioners H Appendix H to Part 305 Commercial Practices FEDERAL TRADE COMMISSION REGULATIONS UNDER SPECIFIC... Manufacturer's rated cooling capacities (Btu's/hr.) Range of SEER's Low High Single Package Units Central...

  7. Definition of a Robust Supervisory Control Scheme for Sodium-Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ponciroli, Roberto; Passerini, Stefano; Vilim, Richard B.

    2016-01-01

    In this work, an innovative control approach for metal-fueled Sodium-cooled Fast Reactors is proposed. With respect to the classical approach adopted for base-load Nuclear Power Plants, an alternative control strategy for operating the reactor at different power levels by respecting the system physical constraints is presented. In order to achieve a higher operational flexibility along with ensuring that the implemented control loops do not influence the system inherent passive safety features, a dedicated supervisory control scheme for the dynamic definition of the corresponding set-points to be supplied to the PID controllers is designed. In particular, the traditional approach based on the adoption of tabulated lookup tables for the set-point definition is found not to be robust enough when failures of the implemented SISO (Single Input Single Output) actuators occur. Therefore, a feedback algorithm based on the Reference Governor approach, which allows for the optimization of reference signals according to the system operating conditions, is proposed.

  8. Coupling of Modular High-Temperature Gas-Cooled Reactor with Supercritical Rankine Cycle

    Directory of Open Access Journals (Sweden)

    Shutang Zhu

    2008-01-01

    Full Text Available This paper presents investigations on the possible combination of modular high-temperature gas-cooled reactor (MHTGR technology with the supercritical (SC steam turbine technology and the prospective deployments of the MHTGR SC power plant. Energy conversion efficiency of steam turbine cycle can be improved by increasing the main steam pressure and temperature. Investigations on SC water reactor (SCWR reveal that the development of SCWR power plants still needs further research and development. The MHTGR SC plant coupling the existing technologies of current MHTGR module design with operation experiences of SC FPP will achieve high cycle efficiency in addition to its inherent safety. The standard once-reheat SC steam turbine cycle and the once-reheat steam cycle with life-steam have been studied and corresponding parameters were computed. Efficiencies of thermodynamic processes of MHTGR SC plants were analyzed, while comparisons were made between an MHTGR SC plant and a designed advanced passive PWR - AP1000. It was shown that the net plant efficiency of an MHTGR SC plant can reach 45% or above, 30% higher than that of AP1000 (35% net efficiency. Furthermore, an MHTGR SC plant has higher environmental competitiveness without emission of greenhouse gases and other pollutants.

  9. Lead-Cooled Fast Reactor Systems and the Fuels and Materials Challenges

    Directory of Open Access Journals (Sweden)

    T. R. Allen

    2007-01-01

    Full Text Available Anticipated developments in the consumer energy market have led developers of nuclear energy concepts to consider how innovations in energy technology can be adapted to meet consumer needs. Properties of molten lead or lead-bismuth alloy coolants in lead-cooled fast reactor (LFR systems offer potential advantages for reactors with passive safety characteristics, modular deployment, and fuel cycle flexibility. In addition to realizing those engineering objectives, the feasibility of such systems will rest on development or selection of fuels and materials suitable for use with corrosive lead or lead-bismuth. Three proposed LFR systems, with varying levels of concept maturity, are described to illustrate their associated fuels and materials challenges. Nitride fuels are generally favored for LFR use over metal or oxide fuels due to their compatibility with molten lead and lead-bismuth, in addition to their high atomic density and thermal conductivity. Ferritic/martensitic stainless steels, perhaps with silicon and/or oxide-dispersion additions for enhanced coolant compatibility and improved high-temperature strength, might prove sufficient for low-to-moderate-temperature LFRs, but it appears that ceramics or refractory metal alloys will be necessary for higher-temperature LFR systems intended for production of hydrogen energy carriers.

  10. High Temperature Fission Chamber for He- and FLiBe-cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bell, Zane W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Giuliano, Dominic R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Holcomb, David Eugene [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Lance, Michael J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Miller, Roger G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Warmack, Robert J. Bruce [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wilson, Dane F. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Mark J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-01-01

    We have evaluated candidate technologies for in-core fission chambers for high-temperature reactors to monitor power level via measurements of neutron flux from start-up through full power at up to 800°C. This research is important because there are no commercially available instruments capable of operating above 550 °C. Component materials and processes were investigated for fission chambers suitable for operation at 800 °C in reactors cooled by molten fluoride salt (FLiBe) or flowing He, with an emphasis placed on sensitivity (≥ 1 cps/nv), service lifetime (2 years at full power), and resistance to direct immersion in FLiBe. The latter gives the instrument the ability to survive accidents involving breach of a thimble. The device is envisioned to be a two-gap, three-electrode instrument constructed from concentric nickel-plated alumina cylinders and using a noble gas–nitrogen fill-gas. We report the results of measurements and calculations of the response of fill gasses, impurity migration in nickel alloy, brazing of the alumina insulator, and thermodynamic calculations.

  11. Procedure of Active Residual Heat Removal after Emergency Shutdown of High-Temperature-Gas-Cooled Reactor

    Directory of Open Access Journals (Sweden)

    Xingtuan Yang

    2014-01-01

    Full Text Available After emergency shutdown of high-temperature-gas-cooled reactor, the residual heat of the reactor core should be removed. As the natural circulation process spends too long period of time to be utilized, an active residual heat removal procedure is needed, which makes use of steam generator and start-up loop. During this procedure, the structure of steam generator may suffer cold/heat shock because of the sudden load of coolant or hot helium at the first few minutes. Transient analysis was carried out based on a one-dimensional mathematical model for steam generator and steam pipe of start-up loop to achieve safety and reliability. The results show that steam generator should be discharged and precooled; otherwise, boiling will arise and introduce a cold shock to the boiling tubes and tube sheet when coolant began to circulate prior to the helium. Additionally, in avoiding heat shock caused by the sudden load of helium, the helium circulation should be restricted to start with an extreme low flow rate; meanwhile, the coolant of steam generator (water should have flow rate as large as possible. Finally, a four-step procedure with precooling process of steam generator was recommended; sensitive study for the main parameters was conducted.

  12. Study of guided wave transmission through complex junction in sodium cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Elie, Q.; Le Bourdais, F.; Jezzine, K.; Baronian, V. [Non Destructive Testing Department at the French Atomic Energy Commission (CEA), Saclay, 91191 Gif sur Yvette CEDEX, (France)

    2015-07-01

    Ultrasonic guided wave techniques are seen as suitable candidates for the inspection of welded structures within sodium cooled fast reactors (SFR), as the long range propagation of guided waves without amplitude attenuation can overcome the accessibility problem due to the liquid sodium. In the context of the development of the Advanced Sodium Test Reactor for Industrial Demonstration (ASTRID), the French Atomic Commission (CEA) investigates non-destructive testing techniques based on guided wave propagation. In this work, guided wave NDT methods are applied to control the integrity of welds located in a junction-type structure welded to the main vessel. The method presented in this paper is based on the analysis of scattering matrices peculiar to each expected defect, and takes advantage of the multi-modal and dispersive characteristics of guided wave generation. In a simulation study, an algorithm developed using the CIVA software is presented. It permits selecting appropriate incident modes to optimize detection and identification of expected flawed configurations. In the second part of this paper, experimental results corresponding to a first validation step of the simulation results are presented. The goal of the experiments is to estimate the effectiveness of the incident mode selection in plates. The results show good agreement between experience and simulation. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1984-07-01

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

  14. Preliminary Study on the High Efficiency Supercritical Pressure Water-Cooled Reactor for Electricity Generation

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Yoon Yeong; Park, Jong Kyun; Cho, Bong Hyun and others

    2006-01-15

    This research has been performed to introduce a concept of supercritical pressure water cooled reactor(SCWR) in Korea The area of research includes core conceptual design, evaluation of candidate fuel, fluid systems conceptual design with mechanical consideration, preparation of safety analysis code, and construction of supercritical pressure heat transfer test facility, SPHINX, and preliminary test. As a result of the research, a set of tools for the reactor core design has been developed and the conceptual core design with solid moderator was proposed. The direct thermodynamic cycle has been studied to find a optimum design. The safety analysis code has also been adapted to supercritical pressure condition. A supercritical pressure CO2 heat transfer test facility has been constructed and preliminary test proved the facility works as expected. The result of this project will be good basis for the participation in the international collaboration under GIF GEN-IV program and next 5-year mid and long term nuclear research program of MOST. The heat transfer test loop, SPHINX, completed as a result of this project may be used for the power cycle study as well as further heat transfer study for the various geometries.

  15. A Conceptual Study on a Supercritical CO2-cooled KAIST Micro Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yu, HwanYeal; Hartanto, Donny; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    Small Modular Reactors (SMRs) are nuclear reactors that are completely built in a factory and shipped to the designated site for installation. As such, the SMR is especially advantageous as a flexible and cost-effective energy source for remote and isolated areas. Furthermore, the concept requires a relatively low capital cost, which makes it attractive for developing countries with limited electricity grid. In addition, the SMR concepts also generate more interest after the Fukushima accident since it can easily be designed with a passive decay heat removal system. One of the major advantages of a water-cooled SMR is its relatively small core size. Nonetheless, in spite of its small core size, the volume and area required for its steam-cycle power conversion unit is still significant. In this study, neutronics feasibility of a fully compact and transportable KAIST micro-modular reactor (MMR) was demonstrated. Rated thermal power of the core was 36.2 MWth with total weight of about 39.6 tons. The core was loaded with 15.5 w/o uranium mono-nitride U15N fuels in order to achieve a targeted lifetime of 20 EFPYs. To achieve targeted lifetime, initial excess reactivity of the core should be quite high, around 4,707 pcm. To reduce the high excess reactivity to about 2,500 pcm, a replaceable burnable absorber was utilized in the design. As a result, the MMR has a 20-year lifetime with a relatively small burnup reactivity swing. Several important safety parameters of the KAIST MMR core were also determined in this study. The Doppler reactivity coefficients and CVRs were demonstrated to negative. Worth of the primary control drums and secondary control rod were much higher than initial excess reactivity.

  16. Sustainability of thorium-uranium in pebble-bed fluoride salt-cooled high temperature reactor

    Directory of Open Access Journals (Sweden)

    Zhu Guifeng

    2016-01-01

    Full Text Available Sustainability of thorium fuel in a Pebble-Bed Fluoride salt-cooled High temperature Reactor (PB-FHR is investigated to find the feasible region of high discharge burnup and negative Flibe (2LiF-BeF2 salt Temperature Reactivity Coefficient (TRC. Dispersion fuel or pellet fuel with SiC cladding and SiC matrix is used to replace the tristructural-isotropic (TRISO coated particle system for increasing fuel loading and decreasing excessive moderation. To analyze the neutronic characteristics, an equilibrium calculation method of thorium fuel self-sustainability is developed. We have compared two refueling schemes (mixing flow pattern and directional flow pattern and two kinds of reflector materials (SiC and graphite. This method found that the feasible region of breeding and negative Flibe TRC is between 20 vol% and 62 vol% fuel loading in the fuel. A discharge burnup could be achieved up to about 200 MWd/kgHM. The case with directional flow pattern and SiC reflector showed superior burnup characteristics but the worst radial power peak factor, while the case with mixing flow pattern and SiC reflector, which was the best tradeoff between discharge burnup and radial power peak factor, could provide burnup of 140 MWd/kgHM and about 1.4 radial power peak factor with 50 vol% dispersion fuel. In addition, Flibe salt displays good neutron properties as a coolant of quasi-fast reactors due to the strong 9Be(n,2n reaction and low neutron absorption of 6Li (even at 1000 ppm in fast spectrum. Preliminary thermal hydraulic calculation shows good safety margin. The greatest challenge of this reactor may be the decades irradiation time of the pebble fuel.

  17. An experimental study of the selective oxidation of ethene in a wall cooled tubular packed bed reactor

    NARCIS (Netherlands)

    Borman, P.C.; Westerterp, K.R.

    1992-01-01

    The selective oxidation of ethene over a silver on ¿-alumina catalyst was studied in a wall cooled tubular reactor. Temperatures were measured inside the bed at different axial and radical positions as well as the overall conversion and selectivity. Locally measured temperatures vary after repacking

  18. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors AGENCY... Commission (NRC) is issuing a new regulatory guide (RG), 1.79.1, ``Initial Test Program of Emergency Core... System (ADAMS): You may access publicly available documents online in the NRC Library at...

  19. Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-3: High Temperature Gas Cooled Reactor Thermal-Hydraulics.

    Science.gov (United States)

    Reihman, Thomas C.

    This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical high temperature gas-cooled reactor (HTGR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module…

  20. CDF modeling of flow and transport processes in the reactor core of a modular high temperature reactor during an air ingress accident; CFD-Modellierung der Stroemungs- und Transportprozesse im Reaktorkern eines modularen Hochtemperaturreaktors waehrend eines Lufteinbruchstoerfalls

    Energy Technology Data Exchange (ETDEWEB)

    Baggemann, Johannes

    2015-05-22

    Generation IV of reactor design is supposed to include inherent safety systems that allow accident management using passive processes (without external energy). The VTR (very high temperature reactor) is graphite moderated with helium cooling. The design concept assumes that in any operational situation the after heat is removed by thermal conduction and radiation. Air ingress is beyond-design accident assuming a leak in the primary circuit triggering oxygen reaction with the hot graphite that could damage the barriers for fission product release. Using 3D CFD (computational fluid dynamics) codes the air ingress scenario is simulated, the flow and transport processes in the reactor core are analyzed. For validation of the modeling heat transport processes were investigated in specific test facilities.

  1. Current design efforts for the gas-cooled fast reactor (GFR)

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, K.D. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, Idaho 83415-3850 (United States)]. e-mail: Kevan.Weaver@inl.gov

    2005-07-01

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

  2. Analysis of the horizontal flow in the advanced gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Duan, Y. [Department of Mechanical Engineering, University of Sheffield, Sheffield S1 3JD (United Kingdom); He, S., E-mail: s.he@sheffield.ac.uk [Department of Mechanical Engineering, University of Sheffield, Sheffield S1 3JD (United Kingdom); Ganesan, P. [Department of Mechanical Engineering, University of Malaya, Kuala Lumpur 50603 (Malaysia); Gotts, J. [EDF Energy, Barnwood, Gloucester GL4 3RS (United Kingdom)

    2014-06-01

    Highlights: • CFD is used to assess the effect of horizontal flows in AGRs. • The horizontal flows can reduce the graphite brick temperature significantly. • Such effects are not taken into consideration in current engineering calculations. • There might be flow instabilities when the fuel channel flow is very low but horizontal flows reduce its possibility. - Abstract: The purpose of the paper is to report a computational investigation of horizontal flows in the UK advanced-gas-cooled reactor (AGR) by using computational fluid dynamics with ANSYS FLUENT. The study is relevant to practical issues encountered in some AGR stations currently in operation in the UK. It is carried out using a comparative approach based on the results of two contrasting models: one simulating the full effect of the cross flow, the other simulating the simplified approach currently employed by the industry which neglects the momentum of the horizontal cross flow. The study reveals that the horizontal cross flow plays a significant role in the cooling of the moderator brick, while the axial variation of the brick geometry also significantly changes the distribution of the temperature within the brick. It is also found that under some circumstances the so-called horizontal inter-brick leakage (HIBL) flow could influence the cooling performance in the narrow gaps, resulting in a local hot spot. Furthermore, there may be flow instabilities in the flows in AGR fuel channels due to the interactions between the flow in the main arrowhead flow passages and that in some narrow passages connected to it, but the influence on the brick temperature is negligible. Horizontal cross flow has an effect of reducing such instabilities.

  3. High-Temperature Gas-Cooled Reactor Technology Development Program: Annual progress report for period ending December 31, 1987

    Energy Technology Data Exchange (ETDEWEB)

    Jones, J.E.,Jr.; Kasten, P.R.; Rittenhouse, P.L.; Sanders, J.P.

    1989-03-01

    The High-Temperature Gas-Cooled Reactor (HTGR) Program being carried out under the US Department of Energy (DOE) continues to emphasize the development of modular high-temperature gas-cooled reactors (MHTGRs) possessing a high degree of inherent safety. The emphasis at this time is to develop the preliminary design of the reference MHTGR and to develop the associated technology base and licensing infrastructure in support of future reactor deployment. A longer-term objective is to realize the full high-temperature potential of HTGRs in gas turbine and high-temperature, process-heat applications. This document summarizes the activities of the HTGR Technology Development Program for the period ending December 31, 1987.

  4. Secondary Cooling Water Quality Management for Multi Purpose Reactor 30 MW GA Siwabessy Indonesia

    Energy Technology Data Exchange (ETDEWEB)

    Sunaryo, Geni Rina, E-mail: genirina@batan.go.i [Center for Reactor Technology and Nuclear Safety (PTRKN-BATAN), Bldg. 80, Puspiptek Area, Serpong, Tangerang 15310 (Indonesia)

    2011-07-01

    Indonesia Multi Purpose Research Reactor (MPR) G.A. Siwabessy 30 MW will be 25 years old in 2011. Series of Non Destructive Test (NDT) were done to understand the current condition such as Eddy Current test for Heat Exchangers, water immersed camera for understanding the tank liner condition, ultrasonic for secondary piping etc. Some deteorization was observed because of ageing and some changing was done. One of them is changing some part of secondary pipe lines because of leaking, with the local ones. For having another 25 years operation life, a proper water quality for secondary cooling water is needed towards corrosion prevention. The main objectives of this experiment is to understand the current water quality of secondary cooling water of RSG-GAS from the aspect of corrosion induced by chemicals and bacteria, and establish procedure for managing the secondary cooling water quality. Methodologies applied are surveillance corrosion by immersing coupon into water observed and followed by visual analyses, corrosion rate determination by electrochemical method with various chemical conditions and total bacteria determination by using test kit. The results show visually that the crevice, galvanic and homogeny corrosion with the current water quality easily be observed for carbon steel represented secondary pipelines at the condition of none oxy bio agent addition. This corrosion is being suppressed by adding the oxy bio agent. The orientation of coupon, vertically and horizontally, gives slightly different effect. The closely corrosion rate was obtained by separately experiment, electrochemical, at the concentration of inhibitor 100ppm is 0.13 {+-} 0.02, which is lower than in the raw water of 0.20 {+-} 0.01 mpy. The total bacteria detected is around 10{sup 7} cfu/ml at none reactor operation and without any anti bacteria added. The oxi bio agent chemical addition suppresses the numbers becomes 10{sup 3} cfu/ml. The SRB bacteria is detected as >10{sup 6} cfu/ml at

  5. Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems

    Energy Technology Data Exchange (ETDEWEB)

    Ferroni, Paolo [Westinghouse Electric Company LLC, Cranberry Township, PA (United States). Global Technology Development; Tatli, Emre [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Czerniak, Luke [Westinghouse Electric Company LLC, Cranberry Township, PA (United States); Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Chien, Hual-Te [Argonne National Lab. (ANL), Argonne, IL (United States); Yoichi, Momozaki [Argonne National Lab. (ANL), Argonne, IL (United States); Bakhtiari, Sasan [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-06-29

    The project “Modeling and Validation of Sodium Plugging for Heat Exchangers in Sodium-cooled Fast Reactor Systems” was conducted jointly by Westinghouse Electric Company (Westinghouse) and Argonne National Laboratory (ANL), over the period October 1, 2013- March 31, 2016. The project’s motivation was the need to provide designers of Sodium Fast Reactors (SFRs) with a validated, state-of-the-art computational tool for the prediction of sodium oxide (Na2O) deposition in small-diameter sodium heat exchanger (HX) channels, such as those in the diffusion bonded HXs proposed for SFRs coupled with a supercritical CO2 (sCO2) Brayton cycle power conversion system. In SFRs, Na2O deposition can potentially occur following accidental air ingress in the intermediate heat transport system (IHTS) sodium and simultaneous failure of the IHTS sodium cold trap. In this scenario, oxygen can travel through the IHTS loop and reach the coldest regions, represented by the cold end of the sodium channels of the HXs, where Na2O precipitation may initiate and continue. In addition to deteriorating HX heat transfer and pressure drop performance, Na2O deposition can lead to channel plugging especially when the size of the sodium channels is small, which is the case for diffusion bonded HXs whose sodium channel hydraulic diameter is generally below 5 mm. Sodium oxide melts at a high temperature well above the sodium melting temperature such that removal of a solid plug such as through dissolution by pure sodium could take a lengthy time. The Sodium Plugging Phenomena Loop (SPPL) was developed at ANL, prior to this project, for investigating Na2O deposition phenomena within sodium channels that are prototypical of the diffusion bonded HX channels envisioned for SFR-sCO2 systems. In this project, a Computational Fluid Dynamic (CFD) model capable of simulating the thermal-hydraulics of the SPPL test

  6. Automatic control system of brain temperature by air-surface cooling for therapeutic hypothermia.

    Science.gov (United States)

    Utsuki, T

    2013-01-01

    An automatic control system of brain temperature by air-surface cooling was developed for therapeutic hypothermia, which is increasingly recommended for hypoxic-ischemic encephalopathy after cardiac arrest and neonatal asphyxia in several guidelines pertinent to resuscitation. Currently, water-surface cooling is the most widespread cooling method in therapeutic hypothermia. However, it requires large electric power for precise control and also needs water-cooling blankets which have potential for compression of patients by its own weight and for water leakage in ICU. Air-surface cooling does not have such problems and is more suitable for clinical use than water-surface cooling, because air has lower specific heat and density as well as the impossibility of the contamination in ICU by its leakage. In the present system, brain temperature of patients is automatically controlled by suitable adjustment of the temperature of the air blowing into the cooling blankets. This adjustment is carried out by the regulation of mixing cool and warm air using proportional control valves. The computer in the developed control apparatus suitably calculates the air temperature and rotation angle of the valves every sampling time on the basis of the optimal-adaptive control algorithm. Thus, the proposed system actualizes automatic control of brain temperature by the inputting only the clinically desired temperature of brain. The control performance of the suggested system was verified by the examination using the mannequin in substitution for an adult patient. In the result, the control error of the head temperature of the mannequin was 0.12 °C on average in spite of the lack of the production capacity of warm air after the re-warming period. Thus, this system serves as a model for the clinically applied system.

  7. Analysis of nocturnal air temperature in districts using mobile measurements and a cooling indicator

    Science.gov (United States)

    Leconte, François; Bouyer, Julien; Claverie, Rémy; Pétrissans, Mathieu

    2016-08-01

    The urban heat island phenomenon is generally defined as an air temperature difference between a city center and the non-urbanized rural areas nearby. However, this description does not encompass the intra-urban temperature differences that exist between neighborhoods in a city. This study investigates the air temperature dynamics of neighborhoods for meteorological conditions that lead to important urban heat island amplitude. Local climate zones (LCZs) have been determined in Nancy, France, and mobile screen-height air temperature measurements are performed using an instrumented vehicle. Initially, hourly measurements are performed within four different LCZs. These results show that air temperature within LCZ demonstrates a nocturnal cooling in two phases, i.e., a first phase between 1 to 3 h before sunset and 3 to 5 h after sunset, and a second phase from 3 to 5 h after sunset to sunrise. During phase 1, neighborhoods exhibit different cooling rate values and air temperature gaps develop between districts, while during phase 2, cooling rates tend to be analogous. Then, a larger meteorological data set is used to investigate these two phases for a selection of 13 LCZs. Normalized cooling rates are calculated between daytime measures and nighttime measures in order to quantify the air temperature dynamics of the studied areas during phase 1. Considering this indicator, three groups are emerging: LCZ compact midrise and open midrise with mean normalized cooling rate values of 0.09 h -1 LCZ large lowrise and open lowrise/sparsely built with mean normalized cooling rate values of 0.011 h -1 LCZ low plants with mean normalized cooling rate values of 0.014 h -1 Results indicate that the relative position of LCZ within the conurbation does not drive air temperature dynamics during phase 1. In addition, measures performed during phase 2 tend to illustrate that cooling rates are similar to all LCZ during this period.

  8. Perceiving nasal patency through mucosal cooling rather than air temperature or nasal resistance.

    Directory of Open Access Journals (Sweden)

    Kai Zhao

    Full Text Available Adequate perception of nasal airflow (i.e., nasal patency is an important consideration for patients with nasal sinus diseases. The perception of a lack of nasal patency becomes the primary symptom that drives these patients to seek medical treatment. However, clinical assessment of nasal patency remains a challenge because we lack objective measurements that correlate well with what patients perceive. The current study examined factors that may influence perceived patency, including air temperature, humidity, mucosal cooling, nasal resistance, and trigeminal sensitivity. Forty-four healthy subjects rated nasal patency while sampling air from three facial exposure boxes that were ventilated with untreated room air, cold air, and dry air, respectively. In all conditions, air temperature and relative humidity inside each box were recorded with sensors connected to a computer. Nasal resistance and minimum airway cross-sectional area (MCA were measured using rhinomanometry and acoustic rhinometry, respectively. General trigeminal sensitivity was assessed through lateralization thresholds to butanol. No significant correlation was found between perceived patency and nasal resistance or MCA. In contrast, air temperature, humidity, and butanol threshold combined significantly contributed to the ratings of patency, with mucosal cooling (heat loss being the most heavily weighted predictor. Air humidity significantly influences perceived patency, suggesting that mucosal cooling rather than air temperature alone provides the trigeminal sensation that results in perception of patency. The dynamic cooling between the airstream and the mucosal wall may be quantified experimentally or computationally and could potentially lead to a new clinical evaluation tool.

  9. Application of ATHLET/DYN3D coupled codes system for fast liquid metal cooled reactor steady state simulation

    Science.gov (United States)

    Ivanov, V.; Samokhin, A.; Danicheva, I.; Khrennikov, N.; Bouscuet, J.; Velkov, K.; Pasichnyk, I.

    2017-01-01

    In this paper the approaches used for developing of the BN-800 reactor test model and for validation of coupled neutron-physic and thermohydraulic calculations are described. Coupled codes ATHLET 3.0 (code for thermohydraulic calculations of reactor transients) and DYN3D (3-dimensional code of neutron kinetics) are used for calculations. The main calculation results of reactor steady state condition are provided. 3-D model used for neutron calculations was developed for start reactor BN-800 load. The homogeneous approach is used for description of reactor assemblies. Along with main simplifications, the main reactor BN-800 core zones are described (LEZ, MEZ, HEZ, MOX, blankets). The 3D neutron physics calculations were provided with 28-group library, which is based on estimated nuclear data ENDF/B-7.0. Neutron SCALE code was used for preparation of group constants. Nodalization hydraulic model has boundary conditions by coolant mass-flow rate for core inlet part, by pressure and enthalpy for core outlet part, which can be chosen depending on reactor state. Core inlet and outlet temperatures were chosen according to reactor nominal state. The coolant mass flow rate profiling through the core is based on reactor power distribution. The test thermohydraulic calculations made with using of developed model showed acceptable results in coolant mass flow rate distribution through the reactor core and in axial temperature and pressure distribution. The developed model will be upgraded in future for different transient analysis in metal-cooled fast reactors of BN type including reactivity transients (control rods withdrawal, stop of the main circulation pump, etc.).

  10. Dehumidifying Air for Cooling & Refrigeration: Nanotechnology Membrane-based Dehumidifier

    Energy Technology Data Exchange (ETDEWEB)

    None

    2010-10-01

    Broad Funding Opportunity Announcement Project: Dais is developing a product called NanoAir which dehumidifies the air entering a building to make air conditioning more energy efficient. The system uses a polymer membrane that allows moisture but not air to pass through it. A vacuum behind the membrane pulls water vapor from the air, and a second set of membranes releases the water vapor outside. The membrane’s high selectivity translates into reduced energy consumption for dehumidification. Dais’ design goals for NanoAir are the use of proprietary materials and processes and industry-standard installation techniques. NanoAir is also complementary to many other energy saving strategies, including energy recovery.

  11. Performance analysis of a combined cycle gas turbine power plant by using various inlet air cooling systems

    Energy Technology Data Exchange (ETDEWEB)

    Rahim, Murad A. [Department of Mechanical Engineering, Gazi University (Turkey)], e-mail: mrahim@gazi.edu.tr

    2011-07-01

    In recent years, the use of gas turbines in combined cycle power plants has increased. Turbine inlet air cooling appears to be the best solution for maximizing both production and efficiency, particularly in a hot climate. The aim of this study is to determine the impact of different air cooling systems on the gas turbine's performance and carbon dioxide emissions. Computer simulations were carried out, using the THERMOFLEX program, on fogging, evaporative cooling, adsorption cooling, and electrical chiller cooling systems as well as on a base case without cooling system. Results showed that inlet air cooling systems are effective in increasing the efficiency of gas turbine power plants. In addition it was found that absorption chillers are the best system for increasing power generation but that economic and source analyses should be conducted before installing a cooling system. This paper demonstrated that inlet air cooling systems have the ability to increase net power generation of gas turbine power plants.

  12. Numerical simulation on internal and external flow field of a SCAL indirect air cooling tower

    Institute of Scientific and Technical Information of China (English)

    TIAN Songfeng; CHAI Yanqin; XIANG Tongqiong; ZHOU Guangsha

    2014-01-01

    According to the actual size of cooling tube bundle and the arrangement of cooling triangle of a surface condenser aluminum exchangers (SCAL)natural draft cooling tower,the geometric model of heat transfer elements at the tower bottom was established.On the basis of the RNG k-εturbulence model and porous medium model,three-dimensional numerical simulation was carried out for the inner and external flow field of the air cooling tower,to investigate the influence of environmental conditions on the tower's operation performance.The results show that,with an increase in ambient wind speed,the inlet air speed at windward side of the tower increases gradually,while that at crosswind side and lee side decreases and tends to be obvious;the tower ventilation rate and outlet air speed increases at first and then decreases,and their maximum values appear when the wind speed is 2 m/s.

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

    Science.gov (United States)

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

    2012-06-01

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

  14. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, September 23, 1976--December 31, 1976

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    This report presents the results of work performed from September 23, 1976 through December 31, 1976 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Process Heat and Direct Cycle Helium Turbine (DCHT) applications, in terms of the affect of simulated reactor primary coolant (impure Helium), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes progress to date on alloy selection for VHTR Nuclear Process Heat (NPH) applications and for DCHT applications. The present status on the simulated reactor helium loop design and on designs for the testing and analysis facilities and equipment is discussed.

  15. Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, July 1, 1979-September 30, 1979

    Energy Technology Data Exchange (ETDEWEB)

    1980-03-07

    The results of work performed from July 1, 1979 through September 30, 1979 on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment. The status of the data management system is presented. In addition, the progress in the screening test program is described.

  16. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, January 1, 1980-March 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-06-25

    Results are presented of work performed on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Included are the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described, including screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850, and 950/sup 0/C.

  17. Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, April 1, 1979-June 30, 1979

    Energy Technology Data Exchange (ETDEWEB)

    1980-01-25

    The results are presented of work performed on the Advanced Gas-Cooled Nuclear Reactor Materials Evaluation and Development Program. The objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment. The status of the data management system is presented. In addition, the progress in the screening test program is described.

  18. Advanced gas cooled nuclear reactor materials evaluation and development program. Progress report, July 1--September 30, 1978

    Energy Technology Data Exchange (ETDEWEB)

    1978-11-24

    Results of work performed from July 1, 1978 through September 30, 1978 on the Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program are presented. Candidate alloys were evaluated for Very High Temperature Reactor Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the affect of simulated reactor primary coolant (Helium containing small amounts of various other gases), the high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. The activities associated with the characterization of the materials for the screening test program are reported, i.e., test specimen preparation, information from the materials characterization tests performed by General Electric, and the status of the simulated reactor helium supply system, testing equipment, and gas chemistry analysis instrumentation and equipment. The status of the data management system is presented.

  19. Application of evaporative cooling on the condenser of window-air-conditioner

    Energy Technology Data Exchange (ETDEWEB)

    Hajidavalloo, Ebrahim [Shahid Chamran University, Department of Mechanical Engineering, Golestan St., Ahwaz, Khoozestan 61355 (Iran, Islamic Republic of)]. E-mail: hajidae_1999@yahoo.com

    2007-08-15

    Reduction of energy consumption is a major concern in the vapor compression refrigeration cycle especially in the area with very hot weather conditions (about 50 deg. C), where window-air-conditioners are usually used to cool homes. In this weather condition performance of air condenser window-air-conditioners decrease sharply and electrical power consumption increase considerably. These problems have activated the research programs in order to improve the performance of window-air-conditioners by enhancing heat transfer rate in the condenser. In this article, a new design with high commercialization potential for incorporating of evaporative cooling in the condenser of window-air-conditioner is introduced and experimentally investigated. A real air conditioner is used to test the innovation by putting two cooling pads in both sides of the air conditioner and injecting water on them in order to cool down the air before it passing over the condenser. The experimental results show that thermodynamic characteristics of new system are considerably improved and power consumption decreases by about 16% and the coefficient of performance increases by about 55%.

  20. Correlation of Cooling Data from an Air-Cooled Cylinder and Several Multicylinder Engines

    Science.gov (United States)

    Pinkel, Benjamin; Ellerbrock, Herman H , Jr

    1940-01-01

    The theory of engine-cylinder cooling developed in a previous report was further substantiated by data obtained on a cylinder from a Wright r-1820-g engine. Equations are presented for the average head and barrel temperatures of this cylinder as functions of the engine and the cooling conditions. These equations are utilized to calculate the variation in cylinder temperature with altitude for level flight and climb. A method is presented for correlating average head and barrel temperatures and temperatures at individual points on the head and the barrel obtained on the test stand and in flight. The method is applied to the correlation and the comparison of data obtained on a number of service engines. Data are presented showing the variation of cylinder temperature with time when the power and the cooling pressure drop are suddenly changed.

  1. Improvement to Air2Air Technology to Reduce Fresh-Water Evaporative Cooling Loss at Coal-Based Thermoelectric Power Plants

    Energy Technology Data Exchange (ETDEWEB)

    Ken Mortensen

    2011-12-31

    This program was undertaken to enhance the manufacturability, constructability, and cost of the Air2Air{TM} Water Conservation and Plume Abatement Cooling Tower, giving a validated cost basis and capability. Air2Air{TM} water conservation technology recovers a portion of the traditional cooling tower evaporate. The Condensing Module provides an air-to-air heat exchanger above the wet fill media, extracting the heat from the hot saturated moist air leaving in the cooling tower and condensing water. The rate of evaporate water recovery is typically 10% - 25% annually, depending on the cooling tower location (climate). This program improved the efficiency and cost of the Air2Air{TM} Water Conservation Cooling Tower capability, and led to the first commercial sale of the product, as described.

  2. Conceptual design for accelerator-driven sodium-cooled sub-critical transmutation reactors using scale laws

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kwang Gu; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-12-31

    The feasibility study on conceptual design methodology for accelerator-driven sodium-cooled sub-critical transmutation reactors has been conducted to optimize the design parameters from the scale laws and validates the reactor performance with the integrated code system. A 1000 MWth sodium-cooled sub-critical transmutation reactor has been scaled and verified through the methodology in this paper, which is referred to Advanced Liquid Metal Reactor (ALMR). A Pb-Bi target material and a partitioned fuel are the liquid phases, and they are cooled by the circulation of secondary Pb-Bi coolant and by primary sodium coolant, respectively. Overall key design parameters are generated from the scale laws and they are improved and validated by the integrated code system. Integrated Code System (ICS) consists of LAHET, HMCNP, ORIGEN2, and COMMIX codes and some files. Through ICS the target region, the core region, and thermal-hydraulic related regions are analyzed once-through Results of conceptual design are attached in this paper. 5 refs., 4 figs., 1 tab. (Author)

  3. TRISO-Coated Fuel Processing to Support High Temperature Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Del Cul, G.D.

    2002-10-01

    The initial objective of the work described herein was to identify potential methods and technologies needed to disassemble and dissolve graphite-encapsulated, ceramic-coated gas-cooled-reactor spent fuels so that the oxide fuel components can be separated by means of chemical processing. The purpose of this processing is to recover (1) unburned fuel for recycle, (2) long-lived actinides and fission products for transmutation, and (3) other fission products for disposal in acceptable waste forms. Follow-on objectives were to identify and select the most promising candidate flow sheets for experimental evaluation and demonstration and to address the needs to reduce technical risks of the selected technologies. High-temperature gas-cooled reactors (HTGRs) may be deployed in the next -20 years to (1) enable the use of highly efficient gas turbines for producing electricity and (2) provide high-temperature process heat for use in chemical processes, such as the production of hydrogen for use as clean-burning transportation fuel. Also, HTGR fuels are capable of significantly higher burn-up than light-water-reactor (LWR) fuels or fast-reactor (FR) fuels; thus, the HTGR fuels can be used efficiently for transmutation of fissile materials and long-lived actinides and fission products, thereby reducing the inventory of such hazardous and proliferation-prone materials. The ''deep-burn'' concept, described in this report, is an example of this capability. Processing of spent graphite-encapsulated, ceramic-coated fuels presents challenges different from those of processing spent LWR fuels. LWR fuels are processed commercially in Europe and Japan; however, similar infrastructure is not available for processing of the HTGR fuels. Laboratory studies on the processing of HTGR fuels were performed in the United States in the 1960s and 1970s, but no engineering-scale processes were demonstrated. Currently, new regulations concerning emissions will impact the

  4. Numerical model of sprayed air cooled condenser coupled to refrigerating system

    Energy Technology Data Exchange (ETDEWEB)

    Youbi-Idrissi, M.; Macchi-Tejeda, H.; Fournaison, L.; Guilpart, J. [Refrigerating Processes Research Unit, CEMAGEF, Parc de Tourvoie, BP 44, 92163 Antony Cedex (France)

    2007-07-15

    Because of technological, economic and environmental constraints, many refrigeration and air conditioning units are equipped with a simple air cooled condenser. Spraying the condenser seems to be an original solution to improve the energetic performances of such systems. To characterise this energetic benefit, a semi-local mathematical model was developed and applied to a refrigerating machine with and without spraying its air cooled condenser. It is found that, compared to a dry air cooled condenser, both the calorific capacity and machine COP increase by 13% and 55%, respectively. Furthermore, the model shows that a spray flow rate threshold occurs. It should not be exceeded to assure an effective and rational spray use. (author)

  5. Air purification in a reverse-flow reactor: Model simulations vs. experiments

    NARCIS (Netherlands)

    Beld, van de L.; Westerterp, K.R.

    1996-01-01

    The behavior of a reverse-flow reactor was studied for the purification of polluted air by catalytic combustion. A heterogeneous one-dimensional model was extended with a heat balance for the reactor wall. An overall heat transport term is included to account for the small heat losses in radial dire

  6. The Gas-Cooled Fast Reactor: Report on Safety System Design for Decay Heat Removal

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. Marshall; T. Y. C. Wei; E. E. Feldman; M. J. Driscoll; H. Ludewig

    2003-09-01

    The gas-cooled fast reactor (GFR) was chosen as one of the Generation IV nuclear reactor systems to be developed based on its excellent potential for sustainability through reduction of the volume and radiotoxicity of both its own fuel and other spent nuclear fuel, and for extending/utilizing uranium resources orders of magnitude beyond what the current open fuel cycle can realize. In addition, energy conversion at high thermal efficiency is possible with the current designs being considered, thus increasing the economic benefit of the GFR. However, research and development challenges include the ability to use passive decay heat removal systems during accident conditions, survivability of fuels and in-core materials under extreme temperatures and radiation, and economical and efficient fuel cycle processes. This report addresses/discusses the decay heat removal options available to the GFR, and the current solutions. While it is possible to design a GFR with complete passive safety (i.e., reliance solely on conductive and radiative heat transfer for decay heat removal), it has been shown that the low power density results in unacceptable fuel cycle costs for the GFR. However, increasing power density results in higher decay heat rates, and the attendant temperature increase in the fuel and core. Use of active movers, or blowers/fans, is possible during accident conditions, which only requires 3% of nominal flow to remove the decay heat. Unfortunately, this requires reliance on active systems. In order to incorporate passive systems, innovative designs have been studied, and a mix of passive and active systems appears to meet the requirements for decay heat removal during accident conditions.

  7. Analysis of Fluid Flow and Heat Transfer Model for the Pebble Bed High Temperature Gas Cooled Reactor

    Directory of Open Access Journals (Sweden)

    S. Yamoah

    2012-06-01

    Full Text Available The pebble bed type high temperature gas cooled nuclear reactor is a promising option for next generation reactor technology and has the potential to provide high efficiency and cost effective electricity generation. The reactor unit heat transfer poses a challenge due to the complexity associated with the thermalflow design. Therefore to reliably simulate the flow and heat transport of the pebble bed modular reactor necessitates a heat transfer model that deals with radiation as well as thermal convection and conduction. In this study, a model with the capability to simulate fluid flow and heat transfer in the pebble bed modular reactor core has been developed. The developed model was implemented on a personal computer using FORTRAN 95 programming language. Several important fluid flow and heat transfer parameters have been examined: including the pressure drop over the reactor core, the heat transfer coefficient, the Nusselt number and the effective thermal conductivity of the fuel pebbles. Results obtained from the simulation experiments show a uniform pressure in the radial direction for a core to fuel element diameter (D/d ratio>20 and the heat transfer coefficient increases with increasing temperature and coolant mass flow rate. The model can adequately account for the flow and heat transfer phenomenon and the loss of pressure through friction in the pebble bed type high temperature nuclear reactor.

  8. Pre-Conceptual Design of a Fluoride-Salt-Cooled Small Modular Advanced High Temperature Reactor (SmAHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Greene, Sherrell R [ORNL; Gehin, Jess C [ORNL; Holcomb, David Eugene [ORNL; Carbajo, Juan J [ORNL; Ilas, Dan [ORNL; Cisneros, Anselmo T [ORNL; Varma, Venugopal Koikal [ORNL; Corwin, William R [ORNL; Wilson, Dane F [ORNL; Yoder Jr, Graydon L [ORNL; Qualls, A L [ORNL; Peretz, Fred J [ORNL; Flanagan, George F [ORNL; Clayton, Dwight A [ORNL; Bradley, Eric Craig [ORNL; Bell, Gary L [ORNL; Hunn, John D [ORNL; Pappano, Peter J [ORNL; Cetiner, Sacit M [ORNL

    2011-02-01

    This document presents the results of a study conducted at Oak Ridge National Laboratory during 2010 to explore the feasibility of small modular fluoride salt-cooled high temperature reactors (FHRs). A preliminary reactor system concept, SmATHR (for Small modular Advanced High Temperature Reactor) is described, along with an integrated high-temperature thermal energy storage or salt vault system. The SmAHTR is a 125 MWt, integral primary, liquid salt cooled, coated particle-graphite fueled, low-pressure system operating at 700 C. The system employs passive decay heat removal and two-out-of-three , 50% capacity, subsystem redundancy for critical functions. The reactor vessel is sufficiently small to be transportable on standard commercial tractor-trailer transport vehicles. Initial transient analyses indicated the transition from normal reactor operations to passive decay heat removal is accomplished in a manner that preserves robust safety margins at all times during the transient. Numerous trade studies and trade-space considerations are discussed, along with the resultant initial system concept. The current concept is not optimized. Work remains to more completely define the overall system with particular emphasis on refining the final fuel/core configuration, salt vault configuration, and integrated system dynamics and safety behavior.

  9. CLASSIFICATION OF SYSTEMS FOR PASSIVE AFTERHEAT REMOVAL FROM REACTOR CONTAINMENT OF NUCLEAR POWER PLANT WITH WATER-COOLED POWER REACTOR

    Directory of Open Access Journals (Sweden)

    N. Khaled

    2014-01-01

    Full Text Available A classification on systems for passive afterheat removal from reactor containment has been developed in the paper.  The classification permits to make a detailed analysis of various concepts pertaining to systems for passive afterheat removal from reactor containment of new generation. The paper considers main classification features of the given systems.

  10. Numerical Analysis of Magnetic Force of Dry-Type Air-Core Reactor

    Institute of Scientific and Technical Information of China (English)

    LIUZhi-gang; GENGYing-san; WANGJian-hua

    2004-01-01

    This paper presents a coupled magnetic-circuit method for computing the magnetic force of air-core reactor under short-time current. The current and the magnetic flux density are computed first and then the magnetic force is obtained. Thus, the dynamic stability performance of air-core reactor can be analyzed at the design stage to reduce experimental cost and shorten the lead-time of product development.

  11. Optimization of power-cycle arrangements for Supercritical Water cooled Reactors (SCWRs)

    Science.gov (United States)

    Lizon-A-Lugrin, Laure

    The world energy demand is continuously rising due to the increase of both the world population and the standard of life quality. Further, to assure both a healthy world economy as well as adequate social standards, in a relatively short term, new energy-conversion technologies are mandatory. Within this framework, a Generation IV International Forum (GIF) was established by the participation of 10 countries to collaborate for developing nuclear power reactors that will replace the present technology by 2030. The main goals of these nuclear-power reactors are: economic competitiveness, sustainability, safety, reliability and resistance to proliferation. As a member of the GIF, Canada has decided to orient its efforts towards the design of a CANDU-type Super Critical Water-cooled Reactor (SCWR). Such a system must run at a coolant outlet temperature of about 625°C and at a pressure of 25 MPa. It is obvious that at such conditions the overall efficiency of this kind of Nuclear Power Plant (NPP) will compete with actual supercritical water-power boilers. In addition, from a heat-transfer viewpoint, the use of a supercritical fluid allows the limitation imposed by Critical Heat Flux (CHF) conditions, which characterize actual technologies, to be removed. Furthermore, it will be also possible to use direct thermodynamic cycles where the supercritical fluid expands right away in a turbine without the necessity of using intermediate steam generators and/or separators. This work presents several thermodynamic cycles that could be appropriate to run SCWR power plants. Improving both thermal efficiency and mechanical power constitutes a multi-objective optimization problem and requires specific tools. To this aim, an efficient and robust evolutionary algorithm, based on genetic algorithm, is used and coupled to an appropriate power plant thermodynamic simulation model. The results provide numerous combinations to achieve a thermal efficiency higher than 50% with a

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

    Energy Technology Data Exchange (ETDEWEB)

    Marvin, M.D.

    1978-10-31

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

  13. The recent development of fabrication of ODS ferritic steels for supercritical water-cooled reactors core application

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Z.; Li, M.; Liao, L.; Liu, X.; He, P.; Xu, Y.; Chen, W.; Ge, C. [Univ. of Science and Technology Beijing, School of Materials Science and Engineering, Beijing (China)

    2010-07-01

    Development of cladding materials which can work at high temperature is crucial to realize highly efficient and high-burnup operation of Generation IV nuclear energy systems. Oxide dispersion strengthened (ODS) ferritic steel is one of the most promising cladding materials for advanced nuclear reactors, such as supercritical water-cooled reactor. ODS ferritic steels with Cr content of 12, 14 and 18% were designed and fabricated in China through the mechanical alloying (MA) route. The process parameters were discussed and optimized. Mechanical properties were measured at room temperature and high temperature. (author)

  14. Methodology of Supervision by Analysis of Thermal Flux for Thermal Conduction of a Batch Chemical Reactor Equipped with a Monofluid Heating/Cooling System

    Directory of Open Access Journals (Sweden)

    Ghania Henini

    2012-01-01

    Full Text Available We present the thermal behavior of a batch reactor to jacket equipped with a monofluid heating/cooling system. Heating and cooling are provided respectively by an electrical resistance and two plate heat exchangers. The control of the temperature of the reaction is based on the supervision system. This strategy of management of the thermal devices is based on the usage of the thermal flux as manipulated variable. The modulation of the monofluid temperature by acting on the heating power or on the opening degrees of an air-to-open valve that delivers the monofluid to heat exchanger. The study shows that the application of this method for the conduct of the pilot reactor gives good results in simulation and that taking into account the dynamics of the various apparatuses greatly improves ride quality of conduct. In addition thermal control of an exothermic reaction (mononitration shows that the consideration of heat generated in the model representation improve the results by elimination any overshooting of the set-point temperature.

  15. An experimental investigation on air-side performances of finned tube heat exchangers for indirect air-cooling tower

    Directory of Open Access Journals (Sweden)

    Du Xueping

    2014-01-01

    Full Text Available A tremendous quantity of water can be saved if the air cooling system is used, comparing with the ordinary water-cooling technology. In this study, two kinds of finned tube heat exchangers in an indirect air-cooling tower are experimentally studied, which are a plain finned oval-tube heat exchanger and a wavy-finned flat-tube heat exchanger in a cross flow of air. Four different air inlet angles (90°, 60 °, 45°, and 30° are tested separately to obtain the heat transfer and resistance performance. Then the air-side experimental correlations of the Nusselt number and friction factor are acquired. The comprehensive heat transfer performances for two finned tube heat exchangers under four air inlet angles are compared. For the plain finned oval-tube heat exchanger, the vertical angle (90° has the worst performance while 45° and 30° has the best performance at small ReDc and at large ReDc, respectively. For the wavy-finned flat-tube heat exchanger, the worst performance occurred at 60°, while the best performance occurred at 45° and 90° at small ReDc and at large ReDc, respectively. From the comparative results, it can be found that the air inlet angle has completely different effects on the comprehensive heat transfer performance for the heat exchangers with different structures.

  16. Influence of Cooling to Heating Load Ratio on Optimal Supply Water and Air Temperatures in an Air Conditioning System

    Science.gov (United States)

    Karino, Naoki; Shiba, Takashi; Yokoyama, Ryohei; Ito, Koichi

    In planning an air conditioning system, supply water and air temperatures are important factors from the viewpoint of energy saving and cost reduction. For example, lower temperature supply water and air for space cooling reduce the coefficient of performance of a refrigeration machine, and increase the thickness of heat insulation material. However, they enable larger temperature differences, and reduce equipment sizes and power demand. It is also an important subject to evaluate the effect of the supply water and air temperatures on energy saving and cost reduction on the annual basis by considering not only cooling but also heating loads. The purposes of this paper are to propose an optimal planning method for an air conditioning system with large temperature difference, and to analyze the effect of supply water and air temperatures on the long-term economics through a numerical study for an office building. As a result, it is shown that the proposed method effectively determines supply water and air temperatures, and the influence of the cooling to heating load ratio on the long-term economics is clarified.

  17. Safety aspects of fuel behaviour during faults and accidents in pressurised water reactors and in liquid sodium cooled fast breeder reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gittus, J.H. (UKAEA Information Services Branch, London); Matthews, J.R. (UKAEA Harwell Lab. (UK). Theoretical Physics Div.); Potter, P.E. (UKAEA Harwell Lab. (UK). Chemistry Div.)

    1989-07-01

    The good safety record of electrical power generating reactors in the European Community is based on a substantial effort to understand the safety characteristics of the reactors and their fuel. In this paper the present state of knowledge of oxide fuels used in current European reactors is reviewed. The main theme of the paper is the importance of the role of fission products and the chemical state of the fuel on all aspects of fuel behaviour. The paper is split into two parts. The first part deals with those aspects specific to water reactors using UO{sub 2} based fuels. The second part of the paper deals with mixed-oxide fuels and the sodium cooled reactors. In each part the following aspects are described: Chemical constitution of the fuel; fuel performance and failure limits; failed fuel behaviour; fuel behaviour in accidents; and the interactions in degraded cores after hypothetical accidents. Future directions of safety related fuel work in Europe are identified. (orig.).

  18. Thermochemical Analysis of Gas-Cooled Reactor Fuels Containing Am and Pu Oxides

    Energy Technology Data Exchange (ETDEWEB)

    Lindemer, T.B.

    2002-09-05

    Literature values and estimated data for the thermodynamics of the actinide oxides and fission products are applied to explain the chemical behavior in gas-cooled-reactor fuels. Emphasis is placed on the Am-O-C and Pu-O-C systems and the data are used to plot the oxygen chemical potential versus temperature of solid-solid and solid-gas equilibria. These results help explain observations of vaporization in Am oxides, nitrides, and carbides and provide guidance for the ceramic processing of the fuels. The thermodynamic analysis is then extended to the fission product systems and the Si-C-O system. Existing data on oxygen release (primarily as CO) as a function of burnup in the thoria-urania fuel system is reviewed and compared to values calculated from thermodynamic data. The calculations of oxygen release are then extended to the plutonia and americia fuels. Use of ZrC not only as a particle coating that may be more resistant to corrosion by Pd and other noble-metal fission products, but also as a means to getter oxygen released by fission is discussed.

  19. Objective Provision Trees of Reactivity Control Safety Function for Sodium-Cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Bongsuk; Yang, Huichang [TUEV Rheinland Korea Ltd., Seoul (Korea, Republic of); Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2014-05-15

    The purpose of this OPT is first to assure the DiD design during the licensing of Sf, but it will also contribute in evaluating the completeness of regulatory requirements under development by Korea Institute of Nuclear Safety (KINS). Based on the definition of Defense-in-Depth (DiD) levels and safety functions for KALIMER Sodium-Cooled Fast Reactor (SFR), suggested in the reference and, Objective Provision Trees (OPTs) of reactivity control function for level 1, 2, 3 and 4 DiD were developed and suggested in this paper. The challenges and mechanisms and provisions were briefly explained in this paper. Comparing the mechanisms and provisions with the requirements will contribute in identifying the missing requirements. Since the design of Prototype Gen-IV Sf (PGSFR) is not mature yet, the OPT is developed for KALIMER design. Developed level 1 to 4 OPTs in this study can be used for the identification of potential design vulnerabilities. When detailed identification of provisions in terms of design features were achieved through the next step of this study, it can contribute to the establishment of defense-in-depth evaluation frame for the regulatory reviews for the licensing process. In the next stage of this study, other safety function will be researched and findings can be suggested as recommendations for the safety improvement.

  20. Analyses of the reflector tank, cold source, and beam tube cooling for ANS reactor

    Energy Technology Data Exchange (ETDEWEB)

    Marland, S. [Tennessee Univ., Knoxville, TN (United States)

    1992-07-01

    This report describes my work as an intern with Martin Marietta Energy Systems, Inc., in the summer of 1991. I was assigned to the Reactor Technology Engineering Department, working on the Advanced Neutron Source (ANS). My first project was to select and analyze sealing systems for the top of the diverter/reflector tank. This involved investigating various metal seals and calculating the forces necessary to maintain an adequate seal. The force calculations led to an analysis of several bolt patterns and lockring concepts that could be used to maintain a seal on the vessel. Another project involved some pressure vessel stress calculations and the calculation of the center of gravity for the cold source assembly. I also completed some sketches of possible cooling channel patterns for the inner vessel of the cold source. In addition, I worked on some thermal design analyses for the reflector tank and beam tubes, including heat transfer calculations and assisting in Patran and Pthermal analyses. To supplement the ANS work, I worked on other projects. I completed some stress/deflection analyses on several different beams. These analyses were done with the aid of CAASE, a beam-analysis software package. An additional project involved bending analysis on a carbon removal system. This study was done to find the deflection of a complex-shaped beam when loaded with a full waste can.

  1. The value of helium-cooled reactor technologies for transmutation of nuclear waste

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez, C.; Baxter, A. [General Atomics, Los Alamos, NM (United States)

    2001-07-01

    Helium-cooled reactor technologies offer significant advantages in accomplishing the waste transmutation process. They are ideally suited for use with thermal, epithermal, or fast neutron energy spectra. They can provide a relatively hard thermal neutron spectrum for transmutation of fissionable materials such as Pu-239 using ceramic-coated transmutation fuel particles, a graphite moderator, and a non-fertile burnable poison. These features (1) allow deep levels of transmutation with minimal or no intermediate reprocessing, (2) enhance passive decay heat removal via heat conduction and radiation, (3) allow operation at relatively high temperatures for a highly efficient generation of electricity, and (4) discharge the transmuted waste in a form that is highly resistant to corrosion for long times. They also offer the possibility for the use of epithermal neutrons that can interact with transmutable materials more effectively because of the large atomic cross sections in this energy domain. A fast spectrum may be useful for deep burnup of certain minor actinides. For this application, helium is essentially transparent to neutrons, does not degrade neutron energies, and offers the hardest possible neutron energy environment. In this paper, we report results from recent work on materials transmutation balances, safety, value to a geological repository, and economic considerations. (authors)

  2. Gas-cooled fast reactor fuel-cost assessment. Final report, October 1978-September 1979

    Energy Technology Data Exchange (ETDEWEB)

    Thompson, M.L.

    1979-01-01

    This program, contracted to provide a Gas Cooled Fast Reactor (GCFR) fuel assembly fabrication cost assessment, comprised the following basic activities: establish agreement on the ground rules for cost assessment, prepare a fuel factory flow sheet, and prepare a cost assessment for fuel assembly fabrication. Two factory sizes, 250 and 25 MTHM/year, were considered for fuel assembly fabrication cost assessment. The work on this program involved utilizing GE LMFBR cost assessment and fuel factory studies experience to provide a cost assessment of GCFR fuel assembly fabrication. The recent impact of highly sensitive safety and safeguards environment policies on fuel factory containment, safety, quality assurance and safeguards costs are significantly higher than might have been expected just a few years ago. Fuel assembly fabrication costs are significant because they represent an estimated 30 to 60% of the total fuel cycle costs. In light of the relative high cost of fabrication, changes in the core and assembly design may be necessary in order to enhance the overall fuel cycle economics. Fabrication costs are based on similar operations and experience used in other fuel cycle studies. Because of extrapolation of present technology (e.g., remote fuel fabrication versus present contact fabrication) and regulatory requirements, conservative cost estimates were made.

  3. Demonstration of leak-before-break in Japan Sodium cooled Fast Reactor (JSFR) pipes

    Energy Technology Data Exchange (ETDEWEB)

    Wakai, Takashi, E-mail: wakai.takashi@jaea.go.jp [Japan Atomic Energy Agency, 4002 Narita-cho, O-arai, Ibaraki 311 1393 (Japan); Machida, Hideo; Yoshida, Shinji [TEPCO Systems Corporation, 2-37-28 Eitai, Koto-ku, Tokyo 135 0034 (Japan); Xu, Yang [Mitsubishi FBR Systems, Inc., 2-34-17 Jingumae, Shibuya-ku, Tokyo 150 0001 (Japan); Tsukimori, Kazuyuki [Japan Atomic Energy Agency, 4002 Narita-cho, O-arai, Ibaraki 311 1393 (Japan)

    2014-04-01

    This paper describes the leak-before-break (LBB) assessment procedure applicable to Japan Sodium cooled Fast Reactor (JSFR) pipes made of modified 9Cr–1Mo steel. For the sodium pipes of JSFR, the continuous leak monitoring will be adopted as an alternative to a volumetric test of the weld joints under conditions that satisfy LBB. Firstly, a LBB assessment flowchart eliminating uncertainty resulted from small scale leakage, such as self plugging phenomenon and influence of crack surface roughness on leak rate, was proposed. Secondly, a rational unstable fracture assessment technique, taking the compliance changing with crack extension into account, was also proposed. Thirdly, a crack opening displacement (COD) assessment technique was developed, because COD assessment method applicable to JSFR pipes – thin wall and small work hardening material – had not been proposed yet. In addition, fracture toughness tests were performed using compact tension (CT) specimens to obtain the fracture toughness, J{sub IC}, and the crack growth resistance (J–R) curve at elevated temperature. Finally, by using the flowchart, proposed techniques and collected data, LBB assessment for the primary sodium pipes of JSFR was conducted. As a result, LBB aspect was successfully demonstrated with sufficient margins.

  4. Comparative CFD analyses of liquid metal cooled reactor for lunar surface power

    Energy Technology Data Exchange (ETDEWEB)

    Schriener, Timothy M. [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Chemical and Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States); El-Genk, Mohamed S., E-mail: mgenk@unm.edu [Institute for Space and Nuclear Power Studies, University of New Mexico, Albuquerque, NM (United States); Chemical and Nuclear Engineering Department, University of New Mexico, Albuquerque, NM (United States); Mechanical Engineering Department, University of New Mexico, Albuquerque, NM (United States)

    2014-12-15

    Highlights: • Performed 3-D thermal-hydraulics analyses of a NaK-56 cooled, sector core reactor for lunar surface power. • Investigated effects mesh refinements and choice of two-layer k–ε and SST k–ω models on results. • Choice of turbulence model slightly affects calculated temperatures and velocities. • Estimates of pressure losses require a finer mesh for convergence than temperature and velocities fields. • SST k–ω model provides more details of mixing eddies, but requires more computation time. - Abstract: This paper presents the results of comparative CFD and thermal-hydraulics analyses of the Solid Core–Sectored Compact Reactor (SC-SCoRe) for lunar surface power. This fast-neutron spectrum, liquid NaK-56 cooled reactor is loaded with highly enriched UN fuel. It nominally generates 1.0 MW{sub th} for ∼21 full-power years at NaK-56 inlet and exit temperatures of 850 K and 900 K. The analyses examine the realizable two-layer k–ε and the Shear Stress Transport (SST) k–ω turbulence models, with different numerical mesh refinements, for simulating the performance of the SC-SCoRe core during nominal operation as well as investigate the effect on the computation time and calculated parameters. In addition, the results calculated for a single tri-lobe flow channel are compared to those obtained using the Detached Eddy Simulation (DES), a hybrid LES and RANS turbulence model. The numerical mesh refinement beyond 2 × 10{sup 7} cells in the flow channels and the choice of turbulence model slightly affect the calculated fuel, core structure, and liquid metal temperatures. They more strongly affect the pressure losses and the intensity of flow mixing and the formation of turbulence eddies in the inlet and exit plenums and exit duct. With the same numerical mesh refinements, the total computation time with the k–ω model is 40–120% longer than with the k-ε model, while the calculated operation parameters are almost identical. The flow

  5. Performance analysis of a ground-assisted direct evaporative cooling air conditioner

    Energy Technology Data Exchange (ETDEWEB)

    Heidarinejad, Ghassem; Khalajzadeh, Vahid [Department of Mechanical Engineering, Tarbiat Modares University, P.O. Box 14115-143, Tehran (Iran); Delfani, Shahram [Building and Housing Research Center (BHRC), P O Box 13145-1696, Tehran (Iran)

    2010-11-15

    In this paper, the results of performance analysis of a ground-assisted hybrid evaporative cooling system in Tehran have been discussed. A Ground Coupled Circuit (GCC) provides the necessary pre-cooling effects, enabling a Direct Evaporative Cooler (DEC) that cools the air even below its wet-bulb temperature. The GCC includes four vertical ground heat exchangers (GHE) which were arrayed in series configuration. In order to have an accurate prediction of the optimum performance of a GCC, a computational fluid dynamic simulation was performed. Simulation results revealed that the combination of GCC and DEC system could provide comfort condition whereas DEC alone did not. Based on the simulation results the cooling effectiveness of a hybrid system is more than 100%. Thus, this novel hybrid system could decrease the air temperature below the ambient wet-bulb temperature. This environmentally clean and energy efficient system can be considered as an alternative to the mechanical vapor compression systems. (author)

  6. Feasibility analysis of modified AL-6XN steel for structure component application in supercritical water-cooled reactor

    Institute of Scientific and Technical Information of China (English)

    Xinggang LI; Qingzhi YAN; Rong MA; Haoqiang WANG; Changchun GE

    2009-01-01

    Modified AL-6XN austenite steel was patterned after AL-6XN superaustenitic stainless steel by introducing microalloy elements such as zirconium and titanium in order to adapt to recrystallizing thermo-mechanical treatment and further improve crevice corrosion resistance. Modified AL-6XN exhibited comparable tensile strength, and superior plasticity and impact toughness to commercial AL-6XN steel. The effects of aging behavior on corrosion resistance and impact toughness were measured to evaluate the qualification of modified AL-6XN steel as an in-core component and cladding material in a supercritical water-cooled reactor. Attention should be paid to degradation in corrosion resistance and impact toughness after aging for 50 hours when modified AL-6XN steel is considered as one of the candidate materials for in-core components and cladding tubes in supercritical water-cooled reactors.

  7. Air-bubbling, hollow-fiber reactor with cell bleeding and cross-flow filtration.

    Science.gov (United States)

    Nishii, K; Sode, K; Karube, I

    1990-05-01

    Continuous asymmetric reduction of dyhydrooxoisophorone (DOIP) to 4-hydroxy-2,2,6-trimethylcyclo-hexanone (4-HTMCH) was achieved by a thermophilic bacterium Bacillus stearothermophilus NK86-0151. Three reactors were used: an air-bubbling hollow-fiber reactor with cell bleeding and cross-flow filtration, an air-lift reactor, and a CSTR with PAA immobilized cells. The maximum cell concentration of 11.1 g dry wt L(-1) was obtained in an air-bubbling hollow-fiber reactor, while in the other reactors the cell densities were between 3.5 and 4.1 g dry wt L(-1) The optimum bleed ratio was 0.1 at the dilution rate 0.3 h(-1) in the hollow-fiber reactor. The highest viable cell concentration was maintained in the dilution range of 0.4-0.7 h(-1) by a combination of proper cell bleeding and cross-flow filtration. The maximum volumetric productivity of 4-HTMCH reached 826 mg L(-1) h(-1) at the dilution rate 0.54 h(-1). This value was 4 and 2 times higher than those in the air-lift reactor and CSTR, respectively. The increasing viable cell concentration increased the volumetric productivity of 4-HTMCH. A cell free product solution was continuously obtained by cross-flow filtration.

  8. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki [Japan Atomic Energy Agency (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan); Hashimoto, Akihiko; Imai, Yasutomo [NDD Corporation (1-1-6 Jounan, Mito, Ibaraki 310-0803, Japan) (Japan); Ito, Masahiro [NESI Inc. (4002 Narita, O-arai, Ibaraki 311-1393, Japan) (Japan)

    2015-12-31

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  9. Development of numerical simulation system for thermal-hydraulic analysis in fuel assembly of sodium-cooled fast reactor

    Science.gov (United States)

    Ohshima, Hiroyuki; Uwaba, Tomoyuki; Hashimoto, Akihiko; Imai, Yasutomo; Ito, Masahiro

    2015-12-01

    A numerical simulation system, which consists of a deformation analysis program and three kinds of thermal-hydraulics analysis programs, is being developed in Japan Atomic Energy Agency in order to offer methodologies to clarify thermal-hydraulic phenomena in fuel assemblies of sodium-cooled fast reactors under various operating conditions. This paper gives the outline of the system and its applications to fuel assembly analyses as a validation study.

  10. Replacement of Natural Fine Aggregate With Air Cooled Blast Furnace Slag An Industrial By Product

    Directory of Open Access Journals (Sweden)

    Dr. B. Krishna Rao

    2015-07-01

    Full Text Available The aim of the investigation is to replace natural fine aggregatewith Air Cooled Blast Furnace Slag in OPC concrete. At present, nearly million tons of slag is being produced in the steel plants, in India. The generation of slag would be dual problem in disposal difficulty and environmental pollution. Some strategies should be used to utilize the slag effectively. Considering physical properties of metallurgical slags and a series of possibilities for their use in the field of civil constructions, this report demonstrates the possibilities of using air cooled blast furnace slag as partial replacement of sand in concrete. A total of five concrete mixes, containing 0%, 12.5%, 25%, 37.5% and 50% partial replacement of regular sand with air cooled blast furnace slag are investigated in the laboratory. These mixes were tested to determine axial compressive strength, split tensile strength, and flexural strength for 7days, 28days, 56days and 90days.

  11. Experimental Study of Air Conditioning Unit of Evaporative Cooling Assisted Mechanical Refrigeration

    Institute of Scientific and Technical Information of China (English)

    HUANG Xiang; XU Fang-cheng; WU Jun-mei

    2009-01-01

    The evaporative cooling,which assists the refrigeration machinery air-conditioning systems test-rig,has been designed.Its structure and working principle were described,and the performance test was con-ducted and analyzed.The test shows that making full use of the evaporative cooling"free cooling" in Spring and Autumn seasons can fully meet the requirements of air-conditioned comfort through the switch of the function in different seasons.Taking into account the evaporative cooling fan and pump energy consumption,compared with the traditional mechanical refrigeration system,more than 80 percent of energy can be saved,and the ener-gY efficiency ratio of the Unit(EER) is as high as 7.63.Using the two stages of indirect evaporative cooling to pre-cool the new wind in summer,under the conditions of the same air supply temperature requirements,0.83 kg/s chilled water saved can be equivalent to the traditional mechanical refrigeration system,and when the newwind ratio up to 50 percent.more than 10 percent load was reduced in mechanical refrigeration system.The overall EER increased about 35 percent.

  12. Effects of respirator ambient air cooling on thermophysiological responses and comfort sensations.

    Science.gov (United States)

    Caretti, David M; Barker, Daniel J

    2014-01-01

    This investigation assessed the thermophysiological and subjective impacts of different respirator ambient air cooling options while wearing chemical and biological personal protective equipment in a warm environment (32.7 ± 0.4°C, 49.6 ± 6.5% RH). Ten volunteers participated in 90-min heat exposure trials with and without respirator (Control) wear and performed computer-generated tasks while seated. Ambient air cooling was provided to respirators modified to blow air to the forehead (FHC) or to the forehead and the breathing zone (BZC) of a full-facepiece air-purifying respirator using a low-flow (45 L·min(-1)) mini-blower. An unmodified respirator (APR) trial was also completed. The highest body temperatures (TTY) and least favorable comfort ratings were observed for the APR condition. With ambient cooling over the last 60 min of heat exposure, TTY averaged 37.4 ± 0.6°C for Control, 38.0 ± 0.4°C for APR, 37.8 ± 0.5°C for FHC, and 37.6 ± 0.7°C for BZC conditions independent of time. Both the FHC and BZC ambient air cooling conditions reduced facial skin temperatures, reduced the rise in body temperatures, and led to more favorable subjective comfort and thermal sensation ratings over time compared to the APR condition; however statistical differences among conditions were inconsistent. Independent of exposure time, average breathing apparatus comfort scores with BZC (7.2 ± 2.5) were significantly different from both Control (8.9 ± 1.4) and APR (6.5 ± 2.2) conditions when ambient cooling was activated. These findings suggest that low-flow ambient air cooling of the face under low work rate conditions and mild hyperthermia may be a practical method to minimize the thermophysiological strain and reduce perceived respirator discomfort.

  13. Drag and Cooling with Various Forms of Cowling for a "Whirlwind" Radial Air-Cooled Engine I

    Science.gov (United States)

    Weick, Fred E

    1930-01-01

    This report presents the results of an investigation undertaken in the 20-foot Propeller Research Tunnel at Langley Field on the cowling of radial air-cooled engines. A portion of the investigation has been completed, in which several forms and degrees of cowling were tested on Wright "Whirlwind" J-5 engine mounted in the nose of a cabin fuselage. The cowlings varied from the one extreme of an entirely exposed engine to the other in which the engine was entirely inclosed. Cooling tests were made and each cowling modified, if necessary, until the engine cooled approximately as satisfactorily as when it was entirely exposed. Drag tests were then made with each form of cowling, and the effect of the cowling on the propulsive efficiency determined with a metal propeller. The propulsive efficiency was found to be practically the same with all forms of cowling. The drag of the cabin fuselage with uncowled engine was found to be more than three times as great as the drag of the fuselage with engine removed and nose rounded. The conventional forms of cowling, in which at least the tops of the cylinder heads and valve gear are exposed, reduce the drag somewhat, but the cowling entirely covering the engine reduces it 2.6 times as much as the best conventional one. The decrease in drag due to the use of spinners proved to be almost negligible. The use of the cowling completely covering the engine seems entirely practical as regards both cooling and maintenance under service conditions. It must be carefully designed, however, to cool properly. With cabin fuselages its use should result in a substantial increase in high speed over that obtained with present forms of cowling on engines similar in contour to the J-5. (author)

  14. CFD Analysis of Turbulent Flow Phenomena in the Lower Plenum of a Prismatic Gas-Cooled Reactor

    Energy Technology Data Exchange (ETDEWEB)

    T. Gallaway; S.P. Antal; M.Z. Podowski; D.P. Guillen

    2007-09-01

    This paper is concerned with the implementation of a computational model of turbulent flow in a section of the lower plenum of Very High Temperature Reactor (VHTR). The proposed model has been encoded in a state-of-the-art CFD code, NPHASE. The results of NPHASE predictions have been compared against the experimental data collected using a scaled model of a sub-region in the lower plenum of a modular prismatic gas-cooled reactor. It has been shown that the NPHASE-based model is capable of predicting a three-dimensional velocity field in a complex geometrical configuration of VHTR lower plenum. The current and future validations of computational predictions are necessary for design and analysis of new reactor concepts, as well as for safety analysis and licensing calculations.

  15. Study on a method for loading a Li compound to produce tritium using high-temperature gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nakaya, Hiroyuki, E-mail: nakaya@nucl.kyushu-u.ac.jp [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Matsuura, Hideaki [Department of Applied Quantum Physics and Nuclear Engineering, Kyushu University, 744 Motooka, Fukuoka 8190395 (Japan); Katayama, Kazunari [Department of Advanced Energy Engineering Science, Kyushu University, 6-1 Kasuga-koen, Kasuga 8168580 (Japan); Goto, Minoru; Nakagawa, Shigeaki [Japan Atomic Energy Agency, 4002 Oarai, Ibaraki (Japan)

    2015-10-15

    Highlights: • Tritium production by a high-temperature gas-cooled reactor was studied. • The loading method considering tritium outflow suppression was estimated. • A reactor with 600 MWt produced 400–600 g of tritium for 180 days. • A possibility that tritium outflow can be sufficiently suppressed was shown. - Abstract: Tritium production using high-temperature gas-cooled reactors and its outflow from the region loading Li compound into the helium coolant are estimated when considering the suppression of tritium outflow. A Li rod containing a cylindrical Li compound placed in an Al{sub 2}O{sub 3} cladding tube is assumed as a method for loading Li compound. A gas turbine high-temperature reactor of 300 MW electrical nominal capacity (GTHTR300) with 600 MW thermal output power is considered and modeled using the continuous-energy Monte Carlo transport code MVP-BURN, where burn-up simulations are carried out. Tritium outflow is estimated from equilibrium solution for the tritium diffusion equation in the cladding tube. A GTHTR300 can produce 400–600 g of tritium over a 180-day operation using the chosen method of loading the Li compound while minimizing tritium outflow from the cladding tube. Optimizing tritium production while suppressing tritium outflow is discussed.

  16. Preliminary Design Study of Medium Sized Gas Cooled Fast Reactor with Natural Uranium as Fuel Cycle Input

    Science.gov (United States)

    Meriyanti, Su'ud, Zaki; Rijal, K.; Zuhair, Ferhat, A.; Sekimoto, H.

    2010-06-01

    In this study a fesibility design study of medium sized (1000 MWt) gas cooled fast reactors which can utilize natural uranium as fuel cycle input has been conducted. Gas Cooled Fast Reactor (GFR) is among six types of Generation IV Nuclear Power Plants. GFR with its hard neuron spectrum is superior for closed fuel cycle, and its ability to be operated in high temperature (850° C) makes various options of utilizations become possible. To obtain the capability of consuming natural uranium as fuel cycle input, modified CANDLE burn-up scheme[1-6] is adopted this GFR system by dividing the core into 10 parts of equal volume axially. Due to the limitation of thermal hydraulic aspects, the average power density of the proposed design is selected about 70 W/cc. As an optimization results, a design of 1000 MWt reactors which can be operated 10 years without refueling and fuel shuffling and just need natural uranium as fuel cycle input is discussed. The average discharge burn-up is about 280 GWd/ton HM. Enough margin for criticallity was obtained for this reactor.

  17. A 100-kWt NaK-Cooled Space Reactor Concept for an Early-Flight Mission

    Science.gov (United States)

    Poston, David I.

    2003-01-01

    A stainless-steel (SS) sodium-potassium (NaK) cooled reactor could potentially be the first step in utilizing fission technology in space. The sum of all system-level experience for liquid-metal-cooled space reactors has been with NaK, including the SNAP-10a, the only reactor ever launched by the US. This paper describes a 100-kWt NaK reactor, the NaK-100, which is designed to be developed with minimal technical risk. In additional to NaK technology heritage, the NaK-100 uses a proven fuel-form (SS/UO2) and is designed for simplified system integration and testing. The pins are placed within a solid SS prism, and the NaK flows in an annulus between the pins and the prism. The nuclear and thermal-hydraulic performance of the NaK-100 is presented, as well as the major differences between the NaK-100 and SNAP-10a.

  18. Preliminary design package for residential heating/cooling system--Rankine air conditioner redesign

    Energy Technology Data Exchange (ETDEWEB)

    1978-12-01

    This report contains a summary of the preliminary redesign and development of a marketable single-family heating and cooling system. The objectives discussed are the interim design and schedule status of the Residential (3-ton) redesign, problem areas and solutions, and the definition of plans for future design and development activities. The proposed system for a single-family residential heating and cooling system is a single-loop, solar-assisted, hydronic-to-warm-air heating subsystem with solar-assisted domestic water heating and a Rankine-driven expansion air-conditioning subsystem.

  19. Assessment of the Use of Nitrogen Trifluoride for Purifying Coolant and Heat Transfer Salts in the Fluoride Salt-Cooled High-Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Scheele, Randall D.; Casella, Andrew M.

    2010-09-28

    This report provides an assessment of the use of nitrogen trifluoride for removing oxide and water-caused contaminants in the fluoride salts that will be used as coolants in a molten salt cooled reactor.

  20. Gas-cooled reactor programs: high-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1981

    Energy Technology Data Exchange (ETDEWEB)

    1982-06-01

    Information is presented concerning HTGR chemistry; fueled graphite development; irradiation services for General Atomic Company; prestressed concrete pressure vessel development; HTGR structural materials; graphite development; high-temperature reactor physics studies; shielding studies; component flow test loop studies; core support performance test; and application and project assessments.

  1. Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.; Wade, D. C.; Nikiforova, A.; Hanania, P.; Ryu, H. J.; Kulesza, K. P.; Kim, S. J.; Halsey, W. G.; Smith, C. F.; Brown, N. W.; Greenspan, E.; de Caro, M.; Li, N.; Hosemann, P.; Zhang, J.; Yu, H.; Nuclear Engineering Division; LLNL; LANL; Massachusetts Inst. of Tech.; Ecole des Mines de Paris; Oregon State Univ.; Univ.of California at Berkley

    2008-06-23

    This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been made at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics

  2. Performance of introducing outdoor cold air for cooling a plant production system with artificial light

    Directory of Open Access Journals (Sweden)

    Jun eWang

    2016-03-01

    Full Text Available The commercial use of a plant production system with artificial light (PPAL is limited by its high initial construction and operation costs. The electric-energy consumed by heat pumps, applied mainly for cooling, accounts for 15-35% of the total electric-energy used in a PPAL. To reduce the electric-energy consumption, an air exchanger with low capacity (180 W was used for cooling by introducing outdoor cold air. In this experiment, the indoor air temperature in two PPALs (floor area: 6.2 m2 each was maintained at 25ºC and 20ºC during light and dark periods, respectively, for lettuce production. In one PPAL (PPALe, an air exchanger (air flow rate: 250 m3 h-1 was used along with a heat pump (cooling capacity: 3.2 kW to maintain the indoor air temperature at the set-point. The other PPAL (PPALc with only a heat pump (cooling capacity: 3.2 kW was used for reference. Effects of introducing outdoor cold air on energy use efficiency, coefficient of performance (COP, electric-energy consumption for cooling and growth of lettuce were investigated. The results show that: when the air temperature difference between indoor and outdoor ranged from 20.2°C to 30.0°C: 1 the average energy use efficiency of the air exchanger was 2.8 and 3.4 times greater than the COP of the heat pumps in the PPALe and PPALc, respectively; 2 hourly electric-energy consumption in the PPALe reduced by 15.8-73.7% compared with that in the PPALc; 3 daily supply of CO2 in the PPALe reduced from 0.15 kg to 0.04 kg compared with that in the PPALc; 4 no significant difference in lettuce growth was observed in both PPALs. The results indicate that using air exchanger to introduce outdoor cold air should be considered as an effective way to reduce electric-energy consumption for cooling with little effects on plant growth in a PPAL.

  3. Scaling approach and thermal-hydraulic analysis in the reactor cavity cooling system of a high temperature gas -cooled reactor and thermal-jet mixing in a sodium fast reactor

    Science.gov (United States)

    Omotowa, Olumuyiwa A.

    This dissertation develops and demonstrates the application of the top-down and bottom-up scaling methodologies to thermal-hydraulic flows in the reactor cavity cooling system (RCCS) of the high temperature gas reactor (HTGR) and upper plenum of the sodium fast reactor (SFR), respectively. The need to integrate scaled separate effects and integral tests was identified. Experimental studies and computational tools (CFD) have been integrated to guide the engineering design, analysis and assessment of this scaling methods under single and two-phase flow conditions. To test this methods, two applicable case studies are considered, and original contributions are noted. Case 1: "Experimental Study of RCCS for the HTGR". Contributions include validation of scaling analysis using the top-down approach as guide to a ¼-scale integral test facility. System code, RELAP5, was developed based on the derived scaling parameters. Tests performed included system sensitivity to decay heat load and heat sink inventory variations. System behavior under steady-state and transient scenarios were predicted. Results show that the system has the capacity to protect the cavity walls from over-heating during normal operations and provide a means for decay heat removal under accident scenarios. A full width half maximum statistical method was devised to characterize the thermal-hydraulics of the non-linear two-phase oscillatory behavior. This facilitated understanding of the thermal hydraulic coupling of the loop segments of the RCCS, the heat transfer, and the two-phase flashing flow phenomena; thus the impact of scaling overall. Case 2: "Computational Studies of Thermal Jet Mixing in SFR". In the pool-type SFR, susceptible regions to thermal striping are the upper instrumentation structure and the intermediate heat exchanger (IHX). We investigated the thermal mixing above the core to UIS and the potential impact due to poor mixing. The thermal mixing of dual-jet flows at different

  4. High-Altitude Flight Cooling Investigation of a Radial Air-Cooled Engine

    Science.gov (United States)

    1946-08-01

    The charge-air weight flow was measured during flight by venturi meters installed in the two parallel lines between tileintercooler and car?nmetor...pressure was measured by a shrouded total-head tubo installed on a streamline loom on the right wi~ tip. A swiveling static tube, which was calibrated in a

  5. Development of New Air-Cooled Heat Pump Chiller 'Compact Cube'

    Science.gov (United States)

    Ookoshi, Yasushi; Ito, Takuya; Yamaguchi, Hiroshi; Kato, Yohei; Ochiai, Yasutaka; Tanaka, Kosuke; Uji, Yoshihiro; Nakayama, Hiroshi

    Further improvement of the performance is requested to air-cooled heat pump chiller from the viewpoint of the global warming prevention. Smaller unit is needed to facilitate the renewal from absorption chiller to air-cooled heat pump chiller. To meet such needs, we developed compact new air-cooled heat pump chiller with high efficiency, 'Compact cube'. The developed machine is side-flow type with U-shaped fin and tube heat exchangers. With this structure, the uniform air velocity, high packed density of the heat exchangers, and the unit miniaturization have been implemented. The refrigeration cycle with two-evaporating temperature has also been implemented. The cooling COP of this cycle is 2% higher compared with conventional one-evaporating temperature cycle because of the rise of average evaporating temperature. In a new model, a new control system, which controls both capacity of compressors and air flow rate corresponding to heat load, has been implemented. As a result, the developed machine achieved IPLV(Integrated Part load Value) to 6.2(MCHV-P1800AE) which is 29% better than the conventional unit.

  6. Prediction of Air Flow and Temperature Distribution Inside a Yogurt Cooling Room Using Computational Fluid Dynamics

    Directory of Open Access Journals (Sweden)

    A Surendhar

    2015-01-01

    Full Text Available Air flow and heat transfer inside a yogurt cooling room were analysed using Computational Fluid Dynamics. Air flow and heat transfer models were based on 3D, unsteady state, incompressible, Reynolds-averaged Navier-Stokes equations and energy equations. Yogurt cooling room was modelled with the measured geometry using 3D design tool AutoCAD. Yogurt cooling room model was exported into the flow simulation software by specifying properties of inlet air, yogurt, pallet and walls of the room. Packing material was not considered in this study because of less thickness (cup-0.5mm, carton box-1.5mm and negligible resistance created in the conduction of heat. 3D Computational domain was meshed with hexahedral cells and governing equations were solved using explicit finite volume method. Air flow pattern inside the room and the temperature distribution in the bulk of palletized yogurt were predicted. Through validation, the variation in the temperature distribution and velocity vector from the measured value was found to be 2.0oC (maximum and 30% respectively. From the simulation and the measured value of the temperature distribution, it was observed that the temperature was non-uniform over the bulk of yogurt. This might be due to refrigeration capacity, air flow pattern, stacking of yogurt or geometry of the room. Required results were achieved by changing the location of the cooling fan.

  7. Design Report for the ½ Scale Air-Cooled RCCS Tests in the Natural convection Shutdown heat removal Test Facility (NSTF)

    Energy Technology Data Exchange (ETDEWEB)

    Lisowski, D. D. [Argonne National Lab. (ANL), Argonne, IL (United States); Farmer, M. T. [Argonne National Lab. (ANL), Argonne, IL (United States); Lomperski, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Kilsdonk, D. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Bremer, N. [Argonne National Lab. (ANL), Argonne, IL (United States); Aeschlimann, R. W. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-06-01

    The Natural convection Shutdown heat removal Test Facility (NSTF) is a large scale thermal hydraulics test facility that has been built at Argonne National Laboratory (ANL). The facility was constructed in order to carry out highly instrumented experiments that can be used to validate the performance of passive safety systems for advanced reactor designs. The facility has principally been designed for testing of Reactor Cavity Cooling System (RCCS) concepts that rely on natural convection cooling for either air or water-based systems. Standing 25-m in height, the facility is able to supply up to 220 kW at 21 kW/m2 to accurately simulate the heat fluxes at the walls of a reactor pressure vessel. A suite of nearly 400 data acquisition channels, including a sophisticated fiber optic system for high density temperature measurements, guides test operations and provides data to support scaling analysis and modeling efforts. Measurements of system mass flow rate, air and surface temperatures, heat flux, humidity, and pressure differentials, among others; are part of this total generated data set. The following report provides an introduction to the top level-objectives of the program related to passively safe decay heat removal, a detailed description of the engineering specifications, design features, and dimensions of the test facility at Argonne. Specifications of the sensors and their placement on the test facility will be provided, along with a complete channel listing of the data acquisition system.

  8. Reduced-scale water test of natural circulation for decay heat removal in loop-type sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Murakami, T., E-mail: murakami@criepi.denken.or.jp [Central Research Institute of Electric Power Industry, 1646 Abiko, Chiba (Japan); Eguchi, Y., E-mail: eguchi@criepi.denken.or.jp [Central Research Institute of Electric Power Industry, 1646 Abiko, Chiba (Japan); Oyama, K., E-mail: kazuhiro_oyama@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 2-34-17 Jinguumae, Shibuya, Tokyo (Japan); Watanabe, O., E-mail: osamu4_watanabe@mfbr.mhi.co.jp [Mitsubishi FBR Systems, Inc., 2-34-17 Jinguumae, Shibuya, Tokyo (Japan)

    2015-07-15

    Highlights: • The natural circulation characteristics of a loop-type SFR are examined by a water test. • The performance of decay heat removal system is evaluated using a similarity law. • The effects of flow deviation in the parallel piping of a primary loop are clarified. • The reproducibility of the natural circulation test is confirmed. - Abstract: Water tests of a loop-type sodium-cooled fast reactor have been conducted to physically evaluate the natural circulation characteristics. The water test apparatus was manufactured as a 1/10-scale mock-up of the Japan Sodium-Cooled Fast Reactor, which adopts a decay heat removal system (DHRS) utilizing natural circulation. Tests simulating a variety of events and operation conditions clarified the thermal hydraulic characteristics and core-cooling performance of the natural circulation in the primary loop. Operation conditions such as the duration of the pump flow coast-down and the activation time of the DHRS affect the natural circulation characteristics. A long pump flow coast-down cools the upper plenum of the reactor vessel (RV). This causes the loss of the buoyant force in the RV. The test result indicates that a long pump flow coast-down tends to result in a rapid increase in the core temperature because of the loss of the buoyant force. The delayed activation of the DHRS causes a decrease in the natural circulation flow rate and a temperature rise in the RV. Flow rate deviation and a reverse flow appear in the parallel cold-leg piping in some events, which cause thermal stratification in the cold-leg piping. The DHRS prevents the core temperature from fatally rise even for the most severe design-basis event, in which sodium leakage in a secondary loop of the DHRS and the opening failure of a single damper of the air cooler occur simultaneously. In the water test for the case of siphon break in the primary loop, which is one of the design extension conditions, a circulation flow consisting of ascendant

  9. Solid electrolytes for use in lead-bismuth eutectic cooled nuclear reactors

    Science.gov (United States)

    Mariën, A.; Lim, J.; Rosseel, K.; Vandermeulen, W.; Van den Bosch, J.

    2012-08-01

    The operating temperature of electrochemical oxygen sensors can likely be lowered by reducing the solid electrolyte resistance. Most often, these sensors use yttria partially stabilized zirconia (e.g. (Y2O3)0.05(ZrO2)0.95, (5-YSZ)) as solid electrolyte. In this paper, we discuss the use of better conducting ceramics than yttria partially stabilized zirconia, as solid electrolytes for oxygen sensors that can be applied in lead-bismuth eutectic (LBE) cooled nuclear reactors. Two stabilized zirconia ceramics ((Y2O3)0.08(ZrO2)0.92, (8-YSZ); (Sc2O3)0.1(CeO2)0.01(ZrO2)0.89) are investigated as well as a hypostoichiometric perovskite-type La0.8Sr0.2Ga0.8Mg0.2O3-δ. The results of microstructural analyses, thermochemical stability tests in LBE (at 360 °C), as well as mechanical tests and four-probe d.c. conductivity measurements (at 300-800 °C) are discussed and compared with the results that were obtained for a commercially available 5-YSZ (Friatec AG, Germany). Of the three studied ceramics, 8-YSZ was identified as the most promising solid electrolyte to reduce the operating temperature of electrochemical oxygen sensors. http://www.friatec.de/content/friatec/en/Ceramics/FRIALIT-DEGUSSIT-Oxide-Ceramics/downloads/Materials.pdf (this URL was last accessed on February 7th, 2012).

  10. Passive residual energy utilization system in thermal cycles on water-cooled power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Placco, Guilherme M.; Guimaraes, Lamartine N.F., E-mail: placco@ieav.cta.br, E-mail: guimarae@ieav.cta.br [Instituto de Estudos Avancados (IEAV/DCTA) Sao Jose dos Campos, SP (Brazil); Santos, Rubens S. dos, E-mail: rsantos@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN -RJ), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    This work presents a concept of a residual energy utilization in nuclear plants thermal cycles. After taking notice of the causes of the Fukushima nuclear plant accident, an idea arose to adapt a passive thermal circuit as part of the ECCS (Emergency Core Cooling System). One of the research topics of IEAv (Institute for Advanced Studies), as part of the heat conversion of a space nuclear power system is a passive multi fluid turbine. One of the main characteristics of this device is its passive capability of staying inert and be brought to power at moments notice. During the first experiments and testing of this passive device, it became clear that any small amount of gas flow would generate power. Given that in the first stages of the Fukushima accident and even during the whole event there was plenty availability of steam flow that would be the proper condition to make the proposed system to work. This system starts in case of failure of the ECCS, including loss of site power, loss of diesel generators and loss of the battery power. This system does not requires electricity to run and will work with bleed steam. It will generate enough power to supply the plant safety system avoiding overheating of the reactor core produced by the decay heat. This passive system uses a modified Tesla type turbine. With the tests conducted until now, it is possible to ensure that the operation of this new turbine in a thermal cycle is very satisfactory and it performs as expected. (author)

  11. Risk management and maintenance optimization of nuclear reactor cooling piping system

    Science.gov (United States)

    Augé, L.; Capra, B.; Lasne, M.; Bernard, O.; Bénéfice, P.; Comby, R.

    2006-11-01

    Seaside nuclear power plants have to face the ageing of nuclear reactor cooling piping systems. In order to minimize the duration of the production unit shutdown, maintenance operations have to be planned well in advance. In a context where owners of infrastructures tend to extend the life span of their goods while having to keep the safety level maximum, Oxand brings its expertise and know-how in management of infrastructures life cycle. A dedicated methodology relies on several modules that all participate in fixing network optimum replacement dates: expertise on ageing mechanisms (corrosion, cement degradation...) and the associated kinetics, expertise on impacts of ageing on functional integrity of piping systems, predictive simulation based on experience feedback, development of monitoring techniques focused on actual threats. More precisely, Oxand has designed a patented monitoring technique based on optic fiber sensors, which aims at controlling the deterioration level of piping systems. This preventive maintenance enables the owner to determine criteria for network replacement based on degradation impacts. This approach helps the owner justify his maintenance strategy and allows him to demonstrate the management of safety level. More generally, all monitoring techniques used by the owners are developed and coupled to predictive simulation tools, notably thanks to processes based on Bayesian approaches. Methodologies to evaluate and optimize operation budgets, depending on predictions of future functional deterioration and available maintenance solutions are also developed and applied. Finally, all information related to infrastructure ageing and available maintenance options are put together to reach the right solution for safe and performing infrastructure management.

  12. Validation of CONTAIN-LMR code for accident analysis of sodium-cooled fast reactor containments

    Energy Technology Data Exchange (ETDEWEB)

    Gordeev, S.; Hering, W.; Schikorr, M.; Stieglitz, R. [Inst. for Neutron Physic and Reactor Technology, Karlsruhe Inst. of Technology, Campus Nord (Germany)

    2012-07-01

    CONTAIN-LMR 1 is an analytical tool for the containment performance of sodium cooled fast reactors. In this code, the modelling for the sodium fire is included: the oxygen diffusion model for the sodium pool fire, and the liquid droplet model for the sodium spray fire. CONTAIN-LMR is also able to model the interaction of liquid sodium with concrete structure. It may be applicable to different concrete compositions. Testing and validation of these models will help to qualify the simulation results. Three experiments with sodium performed in the FAUNA facility at FZK have been used for the validation of CONTAIN-LMR. For pool fire tests, calculations have been performed with two models. The first model consists of one gas cell representing the volume of the burn compartment. The volume of the second model is subdivided into 32 coupled gas cells. The agreement between calculations and experimental data is acceptable. The detailed pool fire model shows less deviation from experiments. In the spray fire, the direct heating from the sodium burning in the media is dominant. Therefore, single cell modeling is enough to describe the phenomena. Calculation results have reasonable agreement with experimental data. Limitations of the implemented spray model can cause the overestimation of predicted pressure and temperature in the cell atmosphere. The ability of the CONTAIN-LMR to simulate the sodium pool fire accompanied by sodium-concrete reactions was tested using the experimental study of sodium-concrete interactions for construction concrete as well as for shielding concrete. The model provides a reasonably good representation of chemical processes during sodium-concrete interaction. The comparison of time-temperature profiles of sodium and concrete shows, that the model requires modifications for predictions of the test results. (authors)

  13. Conceptual Design of Passive Safety System for Lead-Bismuth Cooled Fast Reactor

    Science.gov (United States)

    Abdullah, A. G.; Nandiyanto, A. B. D.

    2016-04-01

    This paper presents the results of the conceptual design of passive safety systems for reactor power 225 MWth using Pb-Bi coolant. Main purpose of this research is to design of heat removal system from the reactor wall. The heat from the reactor wall is removed by RVACS system using the natural circulation from the atmosphere around the reactor at steady state. The calculation is performed numerically using Newton-Raphson method. The analysis involves the heat transfer systems in a radiation, conduction and natural convection. Heat transfer calculations is performed on the elements of the reactor vessel, outer wall of guard vessel and the separator plate. The simulation results conclude that the conceptual design is able to remove heat 1.33% to 4.67% from the thermal reactor power. It’s can be hypothesized if the reactor had an accident, the system can still overcome the heat due to decay.

  14. Temperature Field Accurate Modeling and Cooling Performance Evaluation of Direct-Drive Outer-Rotor Air-Cooling In-Wheel Motor

    Directory of Open Access Journals (Sweden)

    Feng Chai

    2016-10-01

    Full Text Available High power density outer-rotor motors commonly use water or oil cooling. A reasonable thermal design for outer-rotor air-cooling motors can effectively enhance the power density without the fluid circulating device. Research on the heat dissipation mechanism of an outer-rotor air-cooling motor can provide guidelines for the selection of the suitable cooling mode and the design of the cooling structure. This study investigates the temperature field of the motor through computational fluid dynamics (CFD and presents a method to overcome the difficulties in building an accurate temperature field model. The proposed method mainly includes two aspects: a new method for calculating the equivalent thermal conductivity (ETC of the air-gap in the laminar state and an equivalent treatment to the thermal circuit that comprises a hub, shaft, and bearings. Using an outer-rotor air-cooling in-wheel motor as an example, the temperature field of this motor is calculated numerically using the proposed method; the results are experimentally verified. The heat transfer rate (HTR of each cooling path is obtained using the numerical results and analytic formulas. The influences of the structural parameters on temperature increases and the HTR of each cooling path are analyzed. Thereafter, the overload capability of the motor is analyzed in various overload conditions.

  15. Using Firn Air for Facility Cooling at the WAIS Divide Site

    Science.gov (United States)

    2014-09-17

    Expedition NICL National Ice core Laboratory NSF National Science Foundation PLR Division of Polar Programs USAP United States Antarctic Program WAIS...ER D C/ CR RE L TR -1 4- 19 Engineering for Polar Operations, Logistics, and Research (EPOLAR) Using Firn Air for Facility Cooling at...acwc.sdp.sirsi.net/client/default. Engineering for Polar Operations, Logistics, and Research (EPOLAR) ERDC/CRREL TR-14-19 September 2014 Using Firn Air for

  16. Study and Evaluation of Innovative Fuel Handling Systems for Sodium-Cooled Fast Reactors: Fuel Handling Route Optimization

    Directory of Open Access Journals (Sweden)

    Franck Dechelette

    2014-01-01

    Full Text Available The research for technological improvement and innovation in sodium-cooled fast reactor is a matter of concern in fuel handling systems in a view to perform a better load factor of the reactor thanks to a quicker fuelling/defueling process. An optimized fuel handling route will also limit its investment cost. In that field, CEA has engaged some innovation study either of complete FHR or on the optimization of some specific components. This paper presents the study of three SFR fuel handling route fully described and compared to a reference FHR option. In those three FHR, two use a gas corridor to transfer spent and fresh fuel assembly and the third uses two casks with a sodium pot to evacuate and load an assembly in parallel. All of them are designed for the ASTRID reactor (1500 MWth but can be extrapolated to power reactors and are compatible with the mutualisation of one FHS coupled with two reactors. These three concepts are then intercompared and evaluated with the reference FHR according to four criteria: performances, risk assessment, investment cost, and qualification time. This analysis reveals that the “mixed way” FHR presents interesting solutions mainly in terms of design simplicity and time reduction. Therefore its study will be pursued for ASTRID as an alternative option.

  17. Gas-cooled reactor programs. Fuel-management positioning and accounting module: FUELMANG Version V1. 11, September 1981

    Energy Technology Data Exchange (ETDEWEB)

    Medlin, T.W.; Hill, K.L.; Johnson, G.L.; Jones, J.E.; Vondy, D.R.

    1982-01-01

    This report documents the code module FUELMANG for fuel management of a reactor. This code may be used to position fuel during the calculation of a reactor history, maintain a mass balance history of the fuel movement, and calculate the unit fuel cycle component of the electrical generation cost. In addition to handling fixed feed fuel without recycle, provision has been made for fuel recycle with various options applied to the recycled fuel. A continuous fueling option is also available with the code. A major edit produced by the code is a detailed summary of the mass balance history of the reactor and a fuel cost analysis of that mass balance history. This code is incorporated in the system containing the VENTURE diffusion theory neutronics code for routine use. Fuel movement according to prescribed instructions is performed without the access of additional user input data during the calculation of a reactor operating history. Local application has been primarily for analysis of the performance of gas-cooled thermal reactor core concepts.

  18. Analysis of radiological accident emissions of a lead-cooled experimental reactor. LEADER Project; Analisis radiologico de las emisiones en caso de accidente de un reactor experimental refrigerado por plomo. Proyecto LEADER

    Energy Technology Data Exchange (ETDEWEB)

    Gomez Salcedo, F.; Cortes Martin, A.

    2013-07-01

    The LEADER project develops a conceptual level industrial size reactor cooled lead and a demonstration plant of this technology. The project objectives are to define the characteristics and design to installation scale reactor using available technologies and short-term components and assess safety aspects conducting a preliminary analysis of the impact of the facility.

  19. Gas-cooled reactor programs. High-temperature gas-cooled reactor technology development program. Annual progress report, December 31, 1983

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Rittenhouse, P.L.; Bartine, D.E.; Sanders, J.P.

    1984-06-01

    ORNL continues to make significant contributions to the national program. In the HTR fuels area, we are providing detailed statistical information on the fission product retention performance of irradiated fuel. Our studies are also providing basic data on the mechanical, physical, and chemical behavior of HTR materials, including metals, ceramics, graphite, and concrete. The ORNL has an important role in the development of improved HTR graphites and in the specification of criteria that need to be met by commercial products. We are also developing improved reactor physics design methods. Our work in component development and testing centers in the Component Flow Test Loop (CFTL), which is being used to evaluate the performance of the HTR core support structure. Other work includes experimental evaluation of the shielding effectiveness of the lower portions of an HTR core. This evaluation is being performed at the ORNL Tower Shielding Facility. Researchers at ORNL are developing welding techniques for attaching steam generator tubing to the tubesheets and are testing ceramic pads on which the core posts rest. They are also performing extensive testing of aggregate materials obtained from potential HTR site areas for possible use in prestressed concrete reactor vessels. During the past year we continued to serve as a peer reviewer of small modular reactor designs being developed by GA and GE with balance-of-plant layouts being developed by Bechtel Group, Inc. We have also evaluated the national need for developing HTRs with emphasis on the longer term applications of the HTRs to fossil conversion processes.

  20. Incorporation of statistical distribution of particle properties in chemical reactor design and operation: the cooled tubular reactor

    NARCIS (Netherlands)

    Wijngaarden, R.J.; Westerterp, K.R.

    1992-01-01

    Pellet heat and mass transfer coefficients inside packed beds do not have definite deterministic values, but are stochastic quantities with a certain distribution. Here, a method is presented to incorporate the stochastic distribution of pellet properties in reactor design and operation models. The

  1. Reactor in search for money. Cooled neutrons for unique research; Reactor zoekt geld. Gekoelde neutronen maken onderzoek mogelijk

    Energy Technology Data Exchange (ETDEWEB)

    Verdult, E.

    2010-12-17

    The modernization of radiation research at the Delft University of Technology depends on subsidies for new instruments. OYSTER (Optimised Yield for Science, Technology and Education of Radiation) is the plan of the Reactor Institute Delft (RID) to realize such a modernization. The article comprises detailed drawings of the inside of the reactor and illustrates the CNIF (Cold Neutron Irradiation Facility) to fight cancer and POSH-PALS (Positron Annihilation Lifetime Spectrometry) to visualize the atomic structure of materials. [Dutch] De modernisering van het stralingsonderzoek aan de Technische Universiteit Delft staat of valt met subsidie voor een nieuwe opzet van het instrumentarium. OYSTER (Optimised Yield for Science, Technology and Education of Radiation) is het plan van het Reactor Instituut Delft (RID) om de installatie te moderniseren. Het artikel bevat gedetailleerde tekeningen van de binnenkant van de reactor en illustreert de CNIF (Cold Neutron Irradiation Facility) voor de bestrijding van kanker en POSH-PALS (Positron Annihilation Lifetime Spectrometry) waarmee de structuur van materialen op atoomniveau inzichtelijk kan worden gemaakt.

  2. Influence of thermal flow field of cooling tower on recirculation ratio of a direct air-cooled system for a power plant

    Institute of Scientific and Technical Information of China (English)

    Zhao Wanli; Liu Peiqing

    2008-01-01

    In order to get thermal flow field of direct air-cooled system,the hot water was supplied to the model of direct air-cooled condenser(ACC).The particle image velocimetery(PIV)experiments were carried out to get thermal flow field of a ACC under different conditions in low velocity wind tunnel,at the same time,the recirculation ratio at cooling tower was measured,so the relationship between flow field characteristics and recirculation ratio of cooling tower can be discussed.From the results we can see that the flow field configuration around cooling tower has great effects on average recirculation ratio under cooling tower.The eddy formed around cooling tower is a key reason that recireulation pro-duces.The eddy intensity relates to velocity magnitude and direction angle,and the configuration of eddy lies on the ge-ometry size of cooling tower.So changing the flow field configuration around cooling tower reasonably can decrease recir-culation ratio under cooling tower,and heat dispel effect of ACC can also be improved.

  3. Comparison of immersed liquid and air cooling of NASA's Airborne Information Management System

    Science.gov (United States)

    Hoadley, A. W.; Porter, A. J.

    1992-01-01

    The Airborne Information Management System (AIMS) is currently under development at NASA Dryden Flight Research Facility. The AIMS is designed as a modular system utilizing surface mounted integrated circuits in a high-density configuration. To maintain the temperature of the integrated circuits within manufacturer's specifications, the modules are to be filled with Fluorinert FC-72. Unlike ground based liquid cooled computers, the extreme range of the ambient pressures experienced by the AIMS requires the FC-72 be contained in a closed system. This forces the latent heat absorbed during the boiling to be released during the condensation that must take within the closed module system. Natural convection and/or pumping carries the heat to the outer surface of the AIMS module where the heat transfers to the ambient air. This paper will present an evaluation of the relative effectiveness of immersed liquid cooling and air cooling of the Airborne Information Management System.

  4. Summary of research and development effort on air and water cooling of gas turbine blades

    Energy Technology Data Exchange (ETDEWEB)

    Fraas, A.P.

    1980-03-01

    The review on air- and water-cooled gas turbines from the 1904 Lemale-Armengaud water-cooled gas turbine, the 1948 to 1952 NACA work, and the program at GE indicates that the potential of air cooling has been largely exploited in reaching temperatures of 1100/sup 0/C (approx. 2000/sup 0/F) in utility service and that further increases in turbine inlet temperature may be obtained with water cooling. The local heat flux in the first-stage turbine rotor with water cooling is very high, yielding high-temperature gradients and severe thermal stresses. Analyses and tests indicate that by employing a blade with an outer cladding of an approx. 1-mm-thick oxidation-resistant high-nickel alloy, a sublayer of a high-thermal-conductivity, high-strength, copper alloy containing closely spaced cooling passages approx. 2 mm in ID to minimize thermal gradients, and a central high-strength alloy structural spar, it appears possible to operate a water-cooled gas turbine with an inlet gas temperature of 1370/sup 0/C. The cooling-water passages must be lined with an iron-chrome-nickel alloy must be bent 90/sup 0/ to extend in a neatly spaced array through the platform at the base of the blade. The complex geometry of the blade design presents truly formidable fabrication problems. The water flow rate to each of many thousands of coolant passages must be metered and held to within rather close limits because the heat flux is so high that a local flow interruption of only a few seconds would lead to a serious failure.Heat losses to the cooling water will run approx. 10% of the heat from the fuel. By recoverying this waste heat for feedwater heating in a command cycle, these heat losses will give a degradation in the power plant output of approx. 5% relative to what might be obtained if no cooling were required. However, the associated power loss is less than half that to be expected with an elegant air cooling system.

  5. Preliminary Analysis on Heat Removal Capacity of Passive Air-Water Combined Cooling Heat Exchanger Using MARS

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seung-Sin; Jeon, Seong-Su; Hong, Soon-Joon [FNC Tech, Yongin (Korea, Republic of); Bae, Sung-Won; Kwon, Tae-Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Current design requirement for working time of PAFS heat exchanger is about 8 hours. Thus, it is not satisfied with the required cooling capability for the long term SBO(Station Black-Out) situation that is required to over 72 hours cooling. Therefore PAFS is needed to change of design for 72 hours cooling. In order to acquirement of long terms cooling using PAFS, heat exchanger tube has to be submerged in water tank for long time. However, water in the tank is evaporated by transferred heat from heat exchanger tubes, so water level is gradually lowered as time goes on. The heat removal capacity of air cooling heat exchanger is core parameter that is used for decision of applicability on passive air-water combined cooling system using PAFS in long term cooling. In this study, the development of MARS input model and plant accident analysis are performed for the prediction of the heat removal capacity of air cooling heat exchanger. From analysis result, it is known that inflow air velocity is the decisive factor of the heat removal capacity and predicted air velocity is lower than required air velocity. But present heat transfer model and predicted air velocity have uncertainty. So, if changed design of PAFS that has over 4.6 kW heat removal capacity in each tube, this type heat exchanger can be applied to long term cooling of the nuclear power plant.

  6. Preferred Air Velocity and Local Cooling Effect of desk fans in warm environments

    DEFF Research Database (Denmark)

    Simone, Angela; Olesen, Bjarne W.

    2013-01-01

    Common experiences, standards, and laboratory studies show that increased air velocity helps to offset warm sensation due to high environmental temperatures. In warm climate regions the opening of windows and the use of desk or ceiling fans are the most common systems to generate increased airflows...... to compensate for higher environmental temperatures at the expense of no or relatively low energy consumption. When using desk fans, local air movement is generated around the occupant and a certain cooling effect is perceived. The impact of the local air movement generated by different air flow patterns....../kg. By a thermal manikin the effect of direct air movement generated by a personal desk fan at 26 °C, 28 °C, or 30 °C room temperatures and the achievable thermal comfort was also analyzed. Results show that it is possible to offset warm sensation within a range of indoor conditions using increased air velocity...

  7. Thermal profile analysis of Doubly-Fed induction generator based wind power converter with air and liquid cooling methods

    DEFF Research Database (Denmark)

    Zhou, Dao; Blaabjerg, Frede; Lau, Mogens;

    2013-01-01

    Today, wind power generation system keeps on moving from onshore to offshore and also upscaling in size. As the lifetime of the wind power converter is prolonged to 20–25 years, this paper will investigate and compare different cooling methods for power modules — the air cooling and the liquid...... cooling seen from a thermal profile assessment point of view. Firstly, an analytical approach from loss profile to thermal profile for the power semiconductor is proposed and verified in a 2 MW Doubly-Fed Induction Generator (DFIG) based wind turbine system. Then, the typical air cooling and liquid...... cooling in wind power converter are analyzed and compared in terms of the mean junction temperature and the junction temperature fluctuation. It is concluded that the liquid cooling approach has a similar junction temperature fluctuation but gives a lower mean junction temperature than the air cooling...

  8. Probabilistic Fracture Mechanics Analysis of Boling Water Reactor Vessel for Cool-Down and Low Temperature Over-Pressurization Transients

    Directory of Open Access Journals (Sweden)

    Jeong Soon Park

    2016-04-01

    Full Text Available The failure probabilities of the reactor pressure vessel (RPV for low temperature over-pressurization (LTOP and cool-down transients are calculated in this study. For the cool-down transient, a pressure–temperature limit curve is generated in accordance with Section XI, Appendix G of the American Society of Mechanical Engineers (ASME code, from which safety margin factors are deliberately removed for the probabilistic fracture mechanics analysis. Then, sensitivity analyses are conducted to understand the effects of some input parameters. For the LTOP transient, the failure of the RPV mostly occurs during the period of the abrupt pressure rise. For the cool-down transient, the decrease of the fracture toughness with temperature and time plays a main role in RPV failure at the end of the cool-down process. As expected, the failure probability increases with increasing fluence, Cu and Ni contents, and initial reference temperature-nil ductility transition (RTNDT. The effect of warm prestressing on the vessel failure probability for LTOP is not significant because most of the failures happen before the stress intensity factor reaches the peak value while its effect reduces the failure probability by more than one order of magnitude for the cool-down transient.

  9. Microbial fouling community analysis of the cooling water system of a nuclear test reactor with emphasis on sulphate reducing bacteria.

    Science.gov (United States)

    Balamurugan, P; Joshi, M Hiren; Rao, T S

    2011-10-01

    Culture and molecular-based techniques were used to characterize bacterial diversity in the cooling water system of a fast breeder test reactor (FBTR). Techniques were selected for special emphasis on sulphate-reducing bacteria (SRB). Water samples from different locations of the FBTR cooling water system, in addition to biofilm scrapings from carbon steel coupons and a control SRB sample were characterized. Whole genome extraction of the water samples and SRB diversity by group specific primers were analysed using nested PCR and denaturing gradient gel electrophoresis (DGGE). The results of the bacterial assay in the cooling water showed that the total culturable bacteria (TCB) ranged from 10(3) to 10(5) cfu ml(-1); iron-reducing bacteria, 10(3) to 10(5) cfu ml(-1); iron oxidizing bacteria, 10(2) to 10(3) cfu ml(-1) and SRB, 2-29 cfu ml(-1). However, the counts of the various bacterial types in the biofilm sample were 2-3 orders of magnitude higher. SRB diversity by the nested PCR-DGGE approach showed the presence of groups 1, 5 and 6 in the FBTR cooling water system; however, groups 2, 3 and 4 were not detected. The study demonstrated that the PCR protocol influenced the results of the diversity analysis. The paper further discusses the microbiota of the cooling water system and its relevance in biofouling.

  10. Air purification in a reverse-flow reactor: Model simulations vs. experiments

    OpenAIRE

    Beld, van de, L.; Westerterp, K.R.

    1996-01-01

    The behavior of a reverse-flow reactor was studied for the purification of polluted air by catalytic combustion. A heterogeneous one-dimensional model was extended with a heat balance for the reactor wall. An overall heat transport term is included to account for the small heat losses in radial direction. The calculations are compared to experimental data without using fit parameters. The agreement between simulations and experiments is generally good. Discrepancies can be explained mainly by...

  11. Use of local convective and radiant cooling at warm environment: effect on thermal comfort and perceived air quality

    DEFF Research Database (Denmark)

    Melikov, Arsen Krikor; Duszyk, Marcin; Krejcirikova, Barbora

    2012-01-01

    The effect of four local cooling devices (convective, radiant and combined) on thermal comfort and perceived air quality reported by 24 subjects at 28 ˚C and 50% RH was studied. The devices studied were: (1) desk cooling fan, (2) personalized ventilation providing clean air, (3) two radiant panels...

  12. Low-power lead-cooled fast reactor loaded with MOX-fuel

    Science.gov (United States)

    Sitdikov, E. R.; Terekhova, A. M.

    2017-01-01

    Fast reactor for the purpose of implementation of research, education of undergraduate and doctoral students in handling innovative fast reactors and training specialists for atomic research centers and nuclear power plants (BRUTs) was considered. Hard neutron spectrum achieved in the fast reactor with compact core and lead coolant. Possibility of prompt neutron runaway of the reactor is excluded due to the low reactivity margin which is less than the effective fraction of delayed neutrons. The possibility of using MOX fuel in the BRUTs reactor was examined. The effect of Keff growth connected with replacement of natural lead coolant to 208Pb coolant was evaluated. The calculations and reactor core model were performed using the Serpent Monte Carlo code.

  13. Design Study of 200MWth Gas Cooled Fast Reactor with Nitride (UN-PuN Fuel Long Life without Refueling

    Directory of Open Access Journals (Sweden)

    Syarifah Ratna Dewi

    2016-01-01

    Full Text Available Design study of 200 MWth Gas Cooled Fast Reactor with UN-PuN fuel long life without refueling has been done. GFR is one type reactor in Generation IV reactor system. It uses helium coolant and fast neutron spectrum. Helium is chemical inert, single phase and low neutron moderation. In this study the calculations are performed by using SRAC code with PIJ calculation for the fuel pin cell calculation and CITATION calculation for core calculation. The data libraries use JENDL 3.2. The variation fuel fractions are 50% until 60%. The diameter active core is 150 cm and the height active core is 100 cm. The reflector radial-axial width is 50 cm. The variation of the powers are 100 MWth up to 500 MWth. The high power causes the high k-eff value. The optimum design is reached when the power is 200 MWth, variation percentage Plutonium for fuel F1:F2:F3=9%:11%:13%. The comparation of fuel:cladding:coolant fraction = 55%:10%:35%. The cooling down time of Plutonium is nine months. The optimum k-eff value is 1.0142 with excess reactivity value 1.403%. The decay of Plutonium decrease k-eff value in the beginning of burn up.

  14. Flow measurement in base cooling air passages of a rotating turbine blade

    Science.gov (United States)

    Liebert, C. H.; Pollack, F. G.

    1974-01-01

    The operational performance is decribed of a shaft-mounted system for measuring the air mass flow rate in the base cooling passages of a rotating turbine blade. Shaft speeds of 0 to 9000 rpm, air mass flow rates of 0.0035 to 0.039 kg/sec (0.0077 to 0.085 lbm/sec), and blade air temperatures of 300 to 385 K (80 to 233 F) were measured. Comparisons of individual rotating blade flows and corresponding stationary supply orifice flows agreed to within 10 percent.

  15. Indoor Air Quality Assessment in a Radiantly Cooled Tropical Building: a Case Study

    Directory of Open Access Journals (Sweden)

    Qi Jie KWONG

    2015-10-01

    Full Text Available Background: Many studies have been conducted to assess the indoor air quality (IAQ of buildings throughout the world because it is closely related to comfort, safety and work productivity of occupants. However, there is still lack of available literature about IAQ in tropical buildings that apply radiant cooling systems in conditioning the indoor air.Methods: This paper reports the results obtained from an IAQ audit that was conducted in a new radiantly cooled building in Malaysia, by focusing on the IAQ and thermal comfort parameters.Results: It was identified that the measured concentration levels for the five indoor air contaminants (CO, CO2, TVOC, formaldehyde and respirable particulates were within the threshold limit values (TLVs specified in the IAQ guidelines. Besides, no significant difference was found between the contaminant levels in each floor of the studied building, and a majority of the respondents did not encounter any form of physical discomfort. There is a risk of condensation problem, judging from the measured RH level.Conclusion: An increase of airflow rate and more dehumidification work in the studied building can be made to improve IAQ and prevention of condensation problem. Nevertheless, these schemes should be implemented carefully to avoid occupants’ discomfort. Relocation of workstations was suggested, especially for the lower floors, which had higher occupancy levels. Keywords: Indoor air quality (IAQ, Radiant cooling systems, IAQ audit, Indoor air contaminants, Condensation 

  16. Scaled Facility Design Approach for Pool-Type Lead-Bismuth Eutectic Cooled Small Modular Reactor Utilizing Natural Circulation

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sangrok; Shin, Yong-Hoon; Lee, Jueun; Hwang, Il Soon [Seoul National University, Seoul (Korea, Republic of)

    2015-10-15

    In low carbon era, nuclear energy is the most prominent energy source of electricity. For steady ecofriendly nuclear energy supply, Generation IV reactors which are future nuclear reactor require safety, sustainability, economics and non-proliferation as four criteria. Lead cooled fast reactor (LFR) is one of these reactor type and Generation IV international forum (GIF) adapted three reference LFR systems which are a small and movable systems with long life without refueling, intermediate size and huge electricity generation system for power grid. NUTRECK (Nuclear Transmutation Energy Center of Korea) has been designed reactor called URANUS (Ubiquitous, Rugged, Accident-forgiving, Non-proliferating, and Ultra-lasting Sustainer) which is small modular reactor and using lead-bismuth eutectic coolant. To prove natural circulation capability of URANUS and analyze design based accidents, scaling mock-up experiment facility will be constructed. In this paper, simple specifications of URANUS will be presented. Then based on this feature, scaling law and scaled facility design results are presented. To validate safety feature and thermodynamics characteristic of URANUS, scaled mockup facility of URANUS is designed based on the scaling law. This mockup adapts two area scale factors, core and lower parts of mock-up are scaled for 3D flow experiment. Upper parts are scaled different size to reduce electricity power and LBE tonnage. This hybrid scaling method could distort some thermal-hydraulic parameters, however, key parameters for experiment will be matched for up-scaling. Detailed design of mock-up will be determined through iteration for design optimization.

  17. Analysis of N-16 concentration in primary cooling system of AP1000 power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rohanda, Anis [Center for Reactor Technology and Nuclear Safety – BATAN Kawasan PUSPIPTEK Gd. No. 80 Serpong, Tangerang Selatan 15310 (Indonesia); Waris, Abdul [Physics Department of ITB, Indonesia anis-rohanda@yahoo.com (Indonesia)

    2015-04-16

    Nitrogen-16 (N-16) is one of the radiation safety parameter on the primary reactor system. The activation product, N-16, is the predominant contributor to the activity in the reactor coolant system during reactor operation. N-16 is activation product derived from activation of O-16 with fast neutron based on {sup 16}O(n,p){sup 16}N reaction. Thus study is needed and it performs to determine N-16 concentration in reactor coolant (primary coolant) in supporting radiation safety. One of the way is using analytical methode based on activation and redecay princip to obtain N-16 concentration. The analysis was performed on the configuration basis and operational of Westinghouse AP1000 power reactor in several monitoring points at coolant reactor system. The results of the calculation of N-16 concentration at the core outlet, reactor vessel outlet, pressurizer line, inlet and outlet of steam generators, primary pumps, reactor vessels inlet and core inlet are: 281, 257, 255, 250, 145, 142, 129 and 112 µCi/gram respectively. The results of analysis compared with AP1000 design control document as standard values. The verification showed very high accuracy comparation between analytical results and standard values.

  18. Safety and core design of large liquid-metal cooled fast breeder reactors

    Science.gov (United States)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

  19. On the safety and performance demonstration tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and validation and verification of computational codes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Bum; Jeong, Ji Young; Lee, Tae Ho; Kim, Sung Kyun; Euh, Dong Jin; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR) has been developed and the validation and verification (V and V) activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1), produced satisfactory results, which were used for the computer codes V and V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs) have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  20. THE INVESTIGATION OF BURNUP CHARACTERISTICS USING THE SERPENT MONTE CARLO CODE FOR A SODIUM COOLED FAST REACTOR

    Directory of Open Access Journals (Sweden)

    MEHMET E. KORKMAZ

    2014-06-01

    Full Text Available In this research, we investigated the burnup characteristics and the conversion of fertile 232Th into fissile 233U in the core of a Sodium-Cooled Fast Reactor (SFR. The SFR fuel assemblies were designed for burning 232Th fuel (fuel pin 1 and 233U fuel (fuel pin 2 and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method and TTA (Transmutation Trajectory Analysis method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff was between 0.964 and 0.954 and peaking factor is 1.88867.

  1. In vessel detection of delayed neutron emitters from clad failure in sodium cooled nuclear reactors: An estimation of the signal

    Science.gov (United States)

    Filliatre, P.; Jammes, C.; Chapoutier, N.; Jeannot, J.-P.; Jadot, F.; Batail, R.; Verrier, D.

    2014-04-01

    The detection of clad failures is mandatory in sodium-cooled fast neutron reactors in compliance with the "clean sodium" concept. An in-vessel detection system, sensitive to delayed neutrons from fission products released into the primary coolant by failures, partially tested in SUPERPHENIX, is foreseen in current SFR projects in order to reduce significantly the delay before an alarm is issued. In this paper, an estimation of the signal received by such a system in case of a failure is derived, taking the French project ASTRID as a working example. This failure induced signal is compared to that of the contribution of the neutrons from the core itself. The sensitivity of the system is defined in terms of minimal detectable surface of clad failure. Possible solutions to improve this sensitivity are discussed, involving either the sensor itself, or the hydraulic design of the vessel in the early stage of the reactor conception.

  2. The investigation of burnup characteristics using the serpent Monte Carlo code for a sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Korkmaz, Mehmet E.; Agar, Osman [Karamanoglu Mehmetbey University, Faculty of Kamil Oezdag Science, Karaman (Turkmenistan)

    2014-06-15

    In this research, we investigated the burnup characteristics and the conversion of fertile {sup 232}Th into fissile {sup 233}U in the core of a Sodium-Cooled Fast Reactor (SFR). The SFR fuel assemblies were designed for burning {sup 232}Th fuel (fuel pin 1) and {sup 233}U fuel (fuel pin 2) and include mixed minor actinide compositions. Monte Carlo simulations were performed using Serpent Code1.1.19 to compare with CRAM (Chebyshev Rational Approximation Method) and TTA (Transmutation Trajectory Analysis) method in the burnup calculation mode. The total heating power generated in the system was assumed to be 2000 MWth. During the reactor operation period of 600 days, the effective multiplication factor (keff) was between 0.964 and 0.954 and peaking factor is 1.88867.

  3. Experimental study of the effects of a transversal air-flow deflector in electronics air-cooling

    Science.gov (United States)

    Rebay, M.; Arfaoui, A.; Padet, J.; Ben Maad, R.

    2011-03-01

    This paper presents an experimental investigation of the influence of a transversal flow deflector on the cooling of a heated block mounted on a flat plate. The deflector is inclined and therefore it guides the air flow to the upper surface of the block. This configuration is simulating the air-cooling of a rectangular integrated circuit or a current converter mounted on an electronic card. The electronic component is assumed dissipating low heat power, as such, air forced convection is still a sufficient cooling way even without fan or heat sink on the component. The measurements are given by hot and cold wires anemometers and by an InfraRed camera. The results give details of the effects of the deflection on the hydrodynamic and the thermal fields on and over the block for different inclination angles. They show that the deviation caused by the deflector may significantly enhance the heat transfer from the component. Deflection is also able to avoid local overheating of the electronic component. Optimum heat transfer rate and homogenised temperature are shown to be obtained with an inclination angle α=30°.

  4. System and method for temperature control in an oxygen transport membrane based reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Sean M.

    2017-02-21

    A system and method for temperature control in an oxygen transport membrane based reactor is provided. The system and method involves introducing a specific quantity of cooling air or trim air in between stages in a multistage oxygen transport membrane based reactor or furnace to maintain generally consistent surface temperatures of the oxygen transport membrane elements and associated reactors. The associated reactors may include reforming reactors, boilers or process gas heaters.

  5. The current status of fluoride salt cooled high temperature reactor (FHR) technology and its overlap with HIF target chamber concepts

    Science.gov (United States)

    Scarlat, Raluca O.; Peterson, Per F.

    2014-01-01

    The fluoride salt cooled high temperature reactor (FHR) is a class of fission reactor designs that use liquid fluoride salt coolant, TRISO coated particle fuel, and graphite moderator. Heavy ion fusion (HIF) can likewise make use of liquid fluoride salts, to create thick or thin liquid layers to protect structures in the target chamber from ablation by target X-rays and damage from fusion neutron irradiation. This presentation summarizes ongoing work in support of design development and safety analysis of FHR systems. Development work for fluoride salt systems with application to both FHR and HIF includes thermal-hydraulic modeling and experimentation, salt chemistry control, tritium management, salt corrosion of metallic alloys, and development of major components (e.g., pumps, heat exchangers) and gas-Brayton cycle power conversion systems. In support of FHR development, a thermal-hydraulic experimental test bay for separate effects (SETs) and integral effect tests (IETs) was built at UC Berkeley, and a second IET facility is under design. The experiments investigate heat transfer and fluid dynamics and they make use of oils as simulant fluids at reduced scale, temperature, and power of the prototypical salt-cooled system. With direct application to HIF, vortex tube flow was investigated in scaled experiments with mineral oil. Liquid jets response to impulse loading was likewise studied using water as a simulant fluid. A set of four workshops engaging industry and national laboratory experts were completed in 2012, with the goal of developing a technology pathway to the design and licensing of a commercial FHR. The pathway will include experimental and modeling efforts at universities and national laboratories, requirements for a component test facility for reliability testing of fluoride salt equipment at prototypical conditions, requirements for an FHR test reactor, and development of a pre-conceptual design for a commercial reactor.

  6. Neutron flux measurements in the side-core region of Hunterston B advanced gas-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Allen, D.A. [Serco, Rutherford House, Quedgeley, Gloucester, GL2 4NF (United Kingdom); Shaw, S.E. [British Energy, Barnett Way, Barnwood, Gloucester, GL4 3RS (United Kingdom); Huggon, A.P.; Steadman, R.J.; Thornton, D.A. [Serco, Rutherford House, Quedgeley, Gloucester, GL2 4NF (United Kingdom); Whiley, G.S. [British Energy, Barnett Way, Barnwood, Gloucester, GL4 3RS (United Kingdom)

    2011-07-01

    The core restraints of advanced gas-cooled reactors are important structural components that are required to maintain the geometric integrity of the cores. A review of neutron dosimetry for the sister stations Hunterston B and Hinkley Point B identified that earlier conservative assessments predicted high thermal neutron dose rates to key components of the restraint structure (the restraint rod welds), with the implication that some of them may be predicted to fail during a seismic event. A revised assessment was therefore undertaken [Thornton, D. A., Allen, D. A., Tyrrell, R. J., Meese, T. C., Huggon, A.P., Whiley, G. S., and Mossop, J. R., 'A Dosimetry Assessment for the Core Restraint of an Advanced Gas Cooled Reactor,' Proceedings of the 13. International Symposium on Reactor Dosimetry (ISRD-13, May 2008), World Scientific, River Edge, NJ, 2009, W. Voorbraak, L. Debarberis, and P. D'hondt, Eds., pp. 679-687] using a detailed 3D model and a Monte Carlo radiation transport program, MCBEND. This reassessment resulted in more realistic fast and thermal neutron dose recommendations, the latter in particular being much lower than had been thought previously. It is now desirable to improve confidence in these predictions by providing direct validation of the MCBEND model through the use of neutron flux measurements. This paper describes the programme of work being undertaken to deploy two neutron flux measurement 'stringers' within the side-core region of one of the Hunterston B reactors for the purpose of validating the MCBEND model. The design of the stringers and the determination of the preferred deployment locations have been informed by the use of detailed MCBEND flux calculations. These computational studies represent a rare opportunity to design a flux measurement beforehand, with the clear intention of minimising the anticipated uncertainties and obtaining measurements that are known to be representative of the neutron fields to which

  7. Keeping cool: use of air conditioning by australians with multiple sclerosis.

    Science.gov (United States)

    Summers, Michael P; Simmons, Rex D; Verikios, George

    2012-01-01

    Despite the known difficulties many people with MS have with high ambient temperatures, there are no reported studies of air conditioning use and MS. This study systematically examined air conditioner use by Australians with MS. A short survey was sent to all participants in the Australian MS Longitudinal Study cohort with a response rate of 76% (n = 2,385). Questions included hours of air-conditioner use, areas cooled, type and age of equipment, and the personal effects of overheating. Air conditioners were used by 81.9% of respondents, with an additional 9.6% who could not afford an air conditioner. Regional and seasonal variation in air conditioning use was reported, with a national annual mean of 1,557 hours running time. 90.7% reported negative effects from overheating including increased fatigue, an increase in other MS symptoms, reduced household and social activities, and reduced work capacity. Households that include people with MS spend between 4 and 12 times more on keeping cool than average Australian households.

  8. Development of dual cooled annular fuel and its possibility to enhance both economy and safety of light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Yanghyun; Kim, Keonsik; Park, Jeongyong; Yang, Yongsik; Kim, Hyungkyu; In, Wangkee; Song, Kunwoo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-03-15

    Over the past few decades, extensive studies have been performed to improve the reliability and safety of light water reactor (LWR) fuel. In recent years, power updating of about 10% is being achieved by modifying safety analysis methodology and subsequent increase in safety margin. But departure from nucleate boiling (DNB) and loss of coolant accident (LOCA) are still two of the most important limiting factors which would restrict power updating more than 10%. Duel cooled annular fuel, cooled in both internal and external cooling channel, has advantages of considerably lower heat flux and lower fuel temperature than conventional solid fuel. While lower heat flus gives higher DNB margin for the same power retie, lower temperature reduces the stored energy of fuel. However, there are many technical issues that should be addressed before any new type of fuel can be considered for application to LWR. This paper describes the key technologies that Korea Atomic Energy Research Institute (KAERI) has developed for dual cooled annular fuel and discusses the feasibility of its application to LWR.

  9. Neutronic/Thermalhydraulic Coupling Technigues for Sodium Cooled Fast Reactor Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Jean Ragusa; Andrew Siegel; Jean-Michel Ruggieri

    2010-09-28

    The objective of this project was to test new coupling algorithms and enable efficient and scalable multi-physics simulations of advanced nuclear reactors, with considerations regarding the implementation of such algorithms in massively parallel environments. Numerical tests were carried out to verify the proposed approach and the examples included some reactor transients. The project was directly related to the Sodium Fast Reactor program element of the Generation IV Nuclear Energy Systems Initiative and the Advanced Fuel cycle Initiative, and, supported the requirement of high-fidelity simulation as a mean of achieving the goals of the presidential Global Nuclear Energy Partnership (GNEP) vision.

  10. Simulation of Radioactive Corrosion Product in Primary Cooling System of Japanese Sodium-Cooled Fast Breeder Reactor

    Science.gov (United States)

    Matuo, Youichirou; Miyahara, Shinya; Izumi, Yoshinobu

    Radioactive Corrosion Product (CP) is a main cause of personal radiation exposure during maintenance with no breached fuel in fast breeder reactor (FBR) plants. The most important CP is 54Mn and 60Co. In order to establish techniques of radiation dose estimation for radiation workers in radiation-controlled areas of the FBR, the PSYCHE (Program SYstem for Corrosion Hazard Evaluation) code was developed. We add the Particle Model to the conventional PSYCHE analytical model. In this paper, we performed calculation of CP transfer in JOYO using an improved calculation code in which the Particle Model was added to the PSYCHE. The C/E (calculated / experimentally observed) value for CP deposition was improved through use of this improved PSYCHE incorporating the Particle Model. Moreover, among the percentage of total radioactive deposition accounted for by CP in particle form, 54Mn was estimated to constitute approximately 20 % and 60Co approximately 40 % in the cold-leg region. These calculation results are consistent with the measured results for the actual cold-leg piping in the JOYO.

  11. Frequency stabilization of internal-mirror He-Ne lasers by air cooling

    Science.gov (United States)

    Qian, Jin; Liu, Zhongyou; Shi, Chunying; Liu, Xiuying; Wang, Jianbo; Yin, Cong; Cai, Shan

    2013-01-01

    Instead of traditional heating method, the cavity length of an internal-mirror He-Ne laser is controlled by air cooling which is generated by a mini cooling fan. A PID servo controlling system is designed to drive the cooling fan tuning the frequency of the laser. The frequency is stabilized by balancing the power of two operating longitudinal modes. Beating with an iodine stabilized He-Ne laser, a relative uncertainty(Δf / ̅f ) of 4.3×10-9 in 5 months, a frequency fluctuation of less than 2.6 MHz and an Allan deviation of 3×10-11 (τ=100 s) in 75 h are obtained.

  12. Review of selected aspects of the Army Gas-Cooled Reactor Systems Program

    Energy Technology Data Exchange (ETDEWEB)

    None

    1965-08-27

    Information is presented concerning the AGCRS program; ML-1 reactor skid refurbishing program; ML-1-IM fabrication status; power conversion system component testing program; ML-1 demonstration test program; and applications of ML-1 technology.

  13. Feasibility Study of Supercritical Light Water Cooled Reactors for Electric Power Production

    Energy Technology Data Exchange (ETDEWEB)

    Philip MacDonald; Jacopo Buongiorno; James Sterbentz; Cliff Davis; Robert Witt; Gary Was; J. McKinley; S. Teysseyre; Luca Oriani; Vefa Kucukboyaci; Lawrence Conway; N. Jonsson: Bin Liu

    2005-02-13

    The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor.

  14. The Use of Compressed Air for Micro-Jet Cooling After MIG Welding

    Directory of Open Access Journals (Sweden)

    Hadryś D.

    2016-09-01

    Full Text Available The material selected for this investigation was low alloy steel weld metal deposit (WMD after MIG welding with micro-jet cooling. The present investigation was aimed as the following tasks: obtained WMD with various amount of acicular ferrite and further analyze impact toughness of WMD in terms of acicular ferrite amount in it. Weld metal deposit (WMD was first time carried out for MIG welding with micro-jet cooling of compressed air and gas mixture of argon and air. Until that moment only argon, helium and nitrogen were tested as micro-jet gases for MIG/MAG processes. An important role in the interpretation of the results can give methods of artificial intelligence.

  15. Natural air convection for cooling of particle accelerator electromagnets: case studies at CERN

    CERN Document Server

    Moreira, Mariana

    Cooling by natural convection is investigated on two air-cooled corrector magnets at CERN. The heat transfer coefficient (HTC) for each magnet is estimated analytically using established empirical correlations for certain geometries. The HTC is also estimated numerically with three-dimensional steady-state finite element simulations. Air convection around the respective coils as well as heat conduction inside the coils are modelled. Different formulas for the HTC are tested in the post-processing of the simulation results. The HTC for each magnet is then determined experimentally by measuring surface temperatures on the coils through time at constant currents. A method to extract the HTC from these temperature curves is developed, in which the curves are fitted to a function that is derived from a thermodynamical analysis of the heating process. Some plausible ranges for the HTC for each magnet for different currents are obtained. The results of the different estimation methods are compared with the experimen...

  16. Cogen-absorption plants for refrigeration purposes and turbine air inlet cooling

    Energy Technology Data Exchange (ETDEWEB)

    Langreck, Juergen [Colibri bv (Netherlands)

    2000-04-01

    Most cogeneration systems produce power and heat but with absorption refrigeration plants (ARP) the products are power and 'cold'. An ARP driven by heat from a turbine exhaust can provide the cooling for the inlet air with very low consumption of electricity, consequently there is a significant increase in power output from the cogeneration unit. Two different ARP systems are currently available but the author describes only the ammonia-water system, which can achieve temperatures down to -60 degrees C. The article discusses the principle behind ARP, the capital cost and returns on investment, how the cogeneration plant is linked to the ARP, ARP for turbine inlet air cooling, and the potential applications of cogeneration-ARP.

  17. Mesoscale climatic simulation of surface air temperature cooling by highly reflective greenhouses in SE Spain.

    Science.gov (United States)

    Campra, Pablo; Millstein, Dev

    2013-01-01

    A long-term local cooling trend in surface air temperature has been monitored at the largest concentration of reflective greenhouses in the world, at the Province of Almeria, SE Spain, associated with a dramatic increase in surface albedo in the area. The availability of reliable long-term climatic field data at this site offers a unique opportunity to test the skill of mesoscale meteorological models describing and predicting the impacts of land use change on local climate. Using the Weather Research and Forecast (WRF) mesoscale model, we have run a sensitivity experiment to simulate the impact of the observed surface albedo change on monthly and annual surface air temperatures. The model output showed a mean annual cooling of 0.25 °C associated with a 0.09 albedo increase, and a reduction of 22.8 W m(-2) of net incoming solar radiation at surface. Mean reduction of summer daily maximum temperatures was 0.49 °C, with the largest single-day decrease equal to 1.3 °C. WRF output was evaluated and compared with observations. A mean annual warm bias (MBE) of 0.42 °C was estimated. High correlation coefficients (R(2) > 0.9) were found between modeled and observed values. This study has particular interest in the assessment of the potential for urban temperature cooling by cool roofs deployment projects, as well as in the evaluation of mesoscale climatic models performance.

  18. Experimental and numerical analysis of air and radiant cooling systems in offices

    DEFF Research Database (Denmark)

    Corgnati, S. P.; Perino, M.; Fracastoro, G. V.;

    2009-01-01

    then extended to the coupled mixing ventilation and cold radiant ceiling panels. In particular, attention was drawn on the evaluation of the main supply jet properties (throw and penetration length) and on the draft risk caused by the cold air drop into the occupied zone. The study shows that such a problem can......This paper analyses office cooling systems based on all air mixing ventilation systems alone or coupled with radiant ceiling panels. This last solution may be effectively applied to retrofit all air systems that are no longer able to maintain a suitable thermal comfort in the indoor environment......, for example in offices with high thermal loads. This study was performed by means of CFD simulations previously validated through an experimental campaign performed in a full scale test room, simulating a typical two-desk office equipped with an all air mixing ventilation system. The numerical studies were...

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

    Science.gov (United States)

    Mella, R.; Wenman, M. R.

    2013-06-01

    Thermo-mechanical contributions to pellet-clad interaction (PCI) in advanced gas-cooled reactors (AGRs) are modelled in the ABAQUS finite element (FE) code. User supplied sub-routines permit the modelling of the non-linear behaviour of AGR fuel through life. Through utilisation of ABAQUS's well-developed pre- and post-processing ability, the behaviour of the axially constrained steel clad fuel was modelled. The 2D axisymmetric model includes thermo-mechanical behaviour of the fuel with time and condition dependent material properties. Pellet cladding gap dynamics and thermal behaviour are also modelled. The model treats heat up as a fully coupled temperature-displacement study. Dwell time and direct power cycling was applied to model the impact of online refuelling, a key feature of the AGR. The model includes the visco-plastic behaviour of the fuel under the stress and irradiation conditions within an AGR core and a non-linear heat transfer model. A multiscale fission gas release model is applied to compute pin pressure; this model is coupled to the PCI gap model through an explicit fission gas inventory code. Whole pin, whole life, models are able to show the impact of the fuel on all segments of cladding including weld end caps and cladding pellet locking mechanisms (unique to AGR fuel). The development of this model in a commercial FE package shows that the development of a potentially verified and future-proof fuel performance code can be created and used. The usability of a FE based fuel performance code would be an enhancement over past codes. Pre- and post-processors have lowered the entry barrier for the development of a fuel performance model to permit the ability to model complicated systems. Typical runtimes for a 5 year axisymmetric model takes less than one hour on a single core workstation. The current model has implemented: Non-linear fuel thermal behaviour, including a complex description of heat flow in the fuel. Coupled with a variety of

  20. Development of Core Heat Removal Objective Provision Trees for Sodium-Cooled Fast Reactor Defense-in-Depth Evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Huichang; Kang, Bongsuk; Lee, Youngho [TUEV Rheinland Korea Ltd., Seoul (Korea, Republic of); Suh, Namduk [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    Based on the definition of Defense-in-Depth levels and safety functions for KALIMER sodium-cooled fast reactor, suggested in the reference and, OPTs for level 1, 2, and 3 defense-in-depth and core heat removal safety function, were developed and suggested in this paper. The purpose of this OPT is first to assure the defensein-depth design during the licensing of Sodium-Cooled Fast Reactors (SFR), but it will also contribute in evaluating the completeness of regulatory requirements under development by Korea Institute of Nuclear Safety (KINS). The challenges and mechanisms and provisions were briefly explained in this paper. Comparing the mechanisms and provisions with the requirements will contribute in identifying the missing requirements. Since the design of PGSFR (Prototype Gen-IV SFR) is not mature yet, the OPT is developed for KALIMER design. Developed OPTs in this study can be used for the identification of potential design vulnerabilities. When detailed identification of provisions in terms of design features were achieved through the next step of this study, it can contribute to the establishment of defensein-depth evaluation frame for the regulatory reviews for the licensing process. At this moment, the identified provisions have both aspects as requirements and design features already adopted in KALIMER design. In the next stage of this study, derived provisions to be adopted will be compared with the actual design features and findings can be suggested as recommendations for the safety improvement.

  1. Forced Convective Air Cooling from Electronic Component Arrays in a Parallel Plate Channel

    Institute of Scientific and Technical Information of China (English)

    D.Y.Cai; Y.P.Gan; 等

    1994-01-01

    This paper discusses air forced convection heat transfer from inline protruding elements arranged in eight rows.The streamwise and spanwise spacings between elements were varied using a splitter plate that can be positioned at three different modular configurations.A set of empirical formulas was presented to correlate the experimental data for the design of air cooling systems.Arrays of componets with one odd-size module have been tested also.Experimental results show that blocks near the entrance and behind the odd-size module have improved performance compared with uniform arrangements.Accordingly,temperature sensitive components are suggested to be arranged in these locations.

  2. Air Monitoring of Emissions from the Fukushima Daiichi Reactor

    Energy Technology Data Exchange (ETDEWEB)

    McNaughton, Michael [Los Alamos National Laboratory; Allen, Shannon P. [Los Alamos National Laboratory; Archuleta, Debra C. [Los Alamos National Laboratory; Brock, Burgandy [Los Alamos National Laboratory; Coronado, Melissa A. [Los Alamos National Laboratory; Dewart, Jean M. [Los Alamos National Laboratory; Eisele, William F. Jr. [Los Alamos National Laboratory; Fuehne, David P. [Los Alamos National Laboratory; Gadd, Milan S. [Los Alamos National Laboratory; Green, Andrew A. [Los Alamos National Laboratory; Lujan, Joan J. [Los Alamos National Laboratory; MacDonell, Carolyn [Los Alamos National Laboratory; Whicker, Jeffrey J. [Los Alamos National Laboratory

    2012-06-12

    In response to the disasters in Japan on March 11, 2011, and the subsequent emissions from Fukushima-Daiichi, we monitored the air near Los Alamos using four air-monitoring systems: the standard AIRNET samplers, the standard rad-NESHAP samplers, the NEWNET system, and high-volume air samplers. Each of these systems has advantages and disadvantages. In combination, they provide a comprehensive set of measurements of airborne radionuclides near Los Alamos during the weeks following March 11. We report air-monitoring measurements of the fission products released from the Fukushima-Daiichi nuclear-power-plant accident in 2011. Clear gamma-spectrometry peaks were observed from Cs-134, Cs-136, Cs-137, I-131, I132, Te-132, and Te-129m. These data, together with measurements of other radionuclides, are adequate for an assessment and assure us that radionuclides from Fukushima Daiichi did not present a threat to human health at or near Los Alamos. The data demonstrate the capabilities of the Los Alamos air-monitoring systems.

  3. Experimental investigation and CFD analysis of a air cooled condenser heat pipe

    Directory of Open Access Journals (Sweden)

    Annamalai Selvan Arul

    2011-01-01

    Full Text Available In the present work a heat pipe of one meter length and 0.031m outer diameter was constructed and the experiments are conducted to determine the surface and vapour temperature at steady and transient conditions for two different input power in the evaporator section and cooling the condenser section by air. A CFD analysis was also carried and the results under steady state conditions are compared with the results obtained from the experiments and reported in this paper.

  4. In-vessel Retention Strategy for High Power Reactors - K-INERI Final Report (includes SBLB Test Results for Task 3 on External Reactor Vessel Cooling (ERVC) Boiling Data and CHF Enhancement Correlations)

    Energy Technology Data Exchange (ETDEWEB)

    F. B. Cheung; J. Yang; M. B. Dizon; J. Rempe

    2005-01-01

    In-vessel retention (IVR) of core melt is a key severe accident management strategy adopted by some operating nuclear power plants and proposed for some advanced light water reactors (ALWRs). If there were inadequate cooling during a reactor accident, a significant amount of core material could become molten and relocate to the lower head of the reactor vessel, as happened in the Three Mile Island Unit 2 (TMI-2) accident. If it is possible to ensure that the vessel head remains intact so that relocated core materials are retained within the vessel, the enhanced safety associated with these plants can reduce concerns about containment failure and associated risk. For example, the enhanced safety of the Westinghouse Advanced 600 MWe PWR (AP600), which relied upon External Reactor Vessel Cooling (ERVC) for IVR, resulted in the U.S. Nuclear Regulatory Commission (US NRC) approving the design without requiring certain conventional features common to existing LWRs. However, it is not clear that currently proposed external reactor vessel cooling (ERVC) without additional enhancements could provide sufficient heat removal for higher-power reactors (up to 1500 MWe). Hence, a collaborative, three-year, U.S. - Korean International Nuclear Energy Research Initiative (INERI) project was completed in which the Idaho National Engineering and Environmental Laboratory (INEEL), Seoul National University (SNU), Pennsylvania State University (PSU), and the Korea Atomic Energy Research Institute (KAERI) investigated the performance of ERVC and an in-vessel core catcher (IVCC) to determine if IVR is feasible for reactors up to 1500 MWe.

  5. Experimental investigation of the influence of the air jet trajectory on convective heat transfer in buildings equipped with air-based and radiant cooling systems

    DEFF Research Database (Denmark)

    Le Dreau, Jerome; Heiselberg, Per; Jensen, Rasmus Lund

    2015-01-01

    -state and dynamic conditions. With the air-based cooling system, a dependency of the convective heat transfer on the air jet trajectory has been observed. New correlations have been developed, introducing a modified Archimedes number to account for the air flow pattern. The accuracy of the new correlations has been...

  6. Sensitivity Analysis of Fuel Centerline Temperatures in SuperCritical Water-cooled Reactors (SCWRs)

    Science.gov (United States)

    Abdalla, Ayman

    SuperCritical Water-cooled Reactors (SCWRs) are one of the six nuclear-reactor concepts currently being developed under the Generation-IV International Forum (GIF). A main advantage of SCW Nuclear Power Plants (NPPs) is that they offer higher thermal efficiencies compared to those of current conventional NPPs. Unlike today's conventional NPPs, which have thermal efficiencies between 30 - 35%, SCW NPPs will have thermal efficiencies within a range of 45 - 50%, owing to high operating temperatures and pressures (i.e., coolant temperatures as high as 625°C at 25 MPa pressure). The use of current fuel bundles with UO2 fuel at the high operating parameters of SCWRs may cause high fuel centerline temperatures, which could lead to fuel failure and fission gas release. Studies have shown that when the Variant-20 (43-element) fuel bundle was examined at SCW conditions, the fuel centerline temperature industry limit of 1850°C for UO2 and the sheath temperature design limit of 850°C might be exceeded. Therefore, new fuel-bundle designs, which comply with the design requirements, are required for future use in SCWRs. The main objective of this study to conduct a sensitivity analysis in order to identify the main factors that leads to fuel centerline temperature reduction. Therefore, a 54-element fuel bundle with smaller diameter of fuel elements compared to that of the 43-element bundle was designed and various nuclear fuels are examined for future use in a generic Pressure Tube (PT) SCWR. The 54-element bundle consists of 53 heated fuel elements with an outer diameter of 9.5 mm and one central unheated element of 20-mm outer diameter which contains burnable poison. The 54-element fuel bundle has an outer diameter of 103.45 mm, which is the same as the outer diameter of the 43-element fuel bundle. After developing the 54-element fuel bundle, one-dimensional heat-transfer analysis was conducted using MATLAB and NIST REFPROP programs. As a result, the Heat Transfer

  7. Performance evaluation of a stack cooling system using CO{sub 2} air conditioner in fuel cell vehicles

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung Chul; Won, Jong Phil [Thermal Management Research Center, Korea Automotive Technology Institute, Chungnam 330-912 (Korea); Park, Yong Sun; Lim, Tae Won [Corporate Research and Development Division, Hyundai-Kia Motors, Gyeonggi 449-912 (Korea); Kim, Min Soo [School of Mechanical and Aerospace Engineering, Seoul National University, Seoul 151-744 (Korea)

    2009-01-15

    A relation between the heat release from a fuel cell stack and an air conditioning system's performance was investigated. The air conditioning system installed in a fuel cell vehicle can be used for stack cooling when additional stack heat release is required over a fixed radiator capacity during high vehicle power generation. This study investigated the performance of a stack cooling system using CO{sub 2} air conditioner at various operating conditions. Also, the heat releasing effectiveness and mutual interference were analyzed and compared with those for the conventional radiator cooling system with/without cabin cooling. When the radiator coolant inlet temperature and flow rate were 65 C and 80 L/min, respectively, for the outdoor air inlet speed of 5 m/s, the heat release of the stack cooling system with the aid of CO{sub 2} air conditioner increased up to 36% more than that of the conventional radiator cooling system with cabin cooling. Furthermore, this increased by 7% versus the case without cabin cooling. (author)

  8. Cooling, freezing and heating with the air cycle: air as the ultimate green refrigerant

    NARCIS (Netherlands)

    Verschoor, M.J.E.

    2000-01-01

    Due to the recent concern about the damage that CFCs cause to the environment (ozone layer, global warming) and the absence of commonly acceptable alternative refrigerants, the search for alternative refrigeration concepts is going on. Air as refrigerant in the Joule-Brayton cycle (air cycle) is one

  9. An Analysis of Methanol and Hydrogen Production via High-Temperature Electrolysis Using the Sodium Cooled Advanced Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shannon M. Bragg-Sitton; Richard D. Boardman; Robert S. Cherry; Wesley R. Deason; Michael G. McKellar

    2014-03-01

    Integration of an advanced, sodium-cooled fast spectrum reactor into nuclear hybrid energy system (NHES) architectures is the focus of the present study. A techno-economic evaluation of several conceptual system designs was performed for the integration of a sodium-cooled Advanced Fast Reactor (AFR) with the electric grid in conjunction with wind-generated electricity. Cases in which excess thermal and electrical energy would be reapportioned within an integrated energy system to a chemical plant are presented. The process applications evaluated include hydrogen production via high temperature steam electrolysis and methanol production via steam methane reforming to produce carbon monoxide and hydrogen which feed a methanol synthesis reactor. Three power cycles were considered for integration with the AFR, including subcritical and supercritical Rankine cycles and a modified supercritical carbon dioxide modified Brayton cycle. The thermal efficiencies of all of the modeled power conversions units were greater than 40%. A thermal efficiency of 42% was adopted in economic studies because two of the cycles either performed at that level or could potentially do so (subcritical Rankine and S-CO2 Brayton). Each of the evaluated hybrid architectures would be technically feasible but would demonstrate a different internal rate of return (IRR) as a function of multiple parameters; all evaluated configurations showed a positive IRR. As expected, integration of an AFR with a chemical plant increases the IRR when “must-take” wind-generated electricity is added to the energy system. Additional dynamic system analyses are recommended to draw detailed conclusions on the feasibility and economic benefits associated with AFR-hybrid energy system operation.

  10. Development of inflatable seals for the rotatable plugs of sodium cooled fast breeder reactors. A review. Pt. I. Key areas

    Energy Technology Data Exchange (ETDEWEB)

    Sinha, Nilay K. [Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam, Tamilnadu (India). Dept. of Atomic Energy (DAE); Raj, Baldev [P.S. Govindaswamy Naidu (PSG) Institutions, Coimbatore, Tamilnadu (India)

    2013-11-15

    Identification of development areas and their implementation for rotatable plug (RP) inflatable seals of Na cooled, 500 Mw (e) Prototype Fast Breeder Reactor (PFBR) and 40 MW (t) Fast Breeder Test Reactor (FBTR) are described, largely based on a late 1990s survey of cover gas seal development (1950s - early 1990s) which defined a set of shortlisted design options and developmental strategy to minimize effort, cost and time. Comparative studies of top shield sealing and evolving FBR designs suggest suitability of inflatable seal as primary barrier in RPs. International experience identified choice and qualification of seal elastomer under synergistic degrading environment of reactor as the prime element of development. The low pressure, non-reinforced, unbeaded, PFBR inflatable seal (made of 50/50 blend of Viton {sup registered} GBL 200S/600S) developed for 10 y life provides a unification scheme for nuclear elastomeric sealing based on 5 peroxide cured fluoroelastomer blend formulations, 1 finite element analysis approach, 1 Teflon-like plasma coating technique and 2 manufacturing processes promising significant gains in standardization, economy and safety. Uniqueness was ab initio development in the absence of established industry or ready-made supply. Part I addresses key areas of design shortlisting, strategy, development and unification with a backdrop of international evolution. (orig.)

  11. The development of a realistic source term for sodium-cooled fast reactors : assessment of current status and future needs.

    Energy Technology Data Exchange (ETDEWEB)

    LaChance, Jeffrey L.; Phillips, Jesse; Parma, Edward J., Jr.; Olivier, Tara Jean; Middleton, Bobby D.

    2011-06-01

    Sodium-cooled fast reactors (SFRs) continue to be proposed and designed throughout the United States and the world. Although the number of SFRs actually operating has declined substantially since the 1980s, a significant interest in advancing these types of reactor systems remains. Of the many issues associated with the development and deployment of SFRs, one of high regulatory importance is the source term to be used in the siting of the reactor. A substantial amount of modeling and experimental work has been performed over the past four decades on accident analysis, sodium coolant behavior, and radionuclide release for SFRs. The objective of this report is to aid in determining the gaps and issues related to the development of a realistic, mechanistically derived source term for SFRs. This report will allow the reader to become familiar with the severe accident source term concept and gain a broad understanding of the current status of the models and experimental work. Further, this report will allow insight into future work, in terms of both model development and experimental validation, which is necessary in order to develop a realistic source term for SFRs.

  12. Investigation of Nuclear Data Libraries with TRIPOLI-4 Monte Carlo Code for Sodium-cooled Fast Reactors

    Science.gov (United States)

    Lee, Y.-K.; Brun, E.

    2014-04-01

    The Sodium-cooled fast neutron reactor ASTRID is currently under design and development in France. Traditional ECCO/ERANOS fast reactor code system used for ASTRID core design calculations relies on multi-group JEFF-3.1.1 data library. To gauge the use of ENDF/B-VII.0 and JEFF-3.1.1 nuclear data libraries in the fast reactor applications, two recent OECD/NEA computational benchmarks specified by Argonne National Laboratory were calculated. Using the continuous-energy TRIPOLI-4 Monte Carlo transport code, both ABR-1000 MWth MOX core and metallic (U-Pu) core were investigated. Under two different fast neutron spectra and two data libraries, ENDF/B-VII.0 and JEFF-3.1.1, reactivity impact studies were performed. Using JEFF-3.1.1 library under the BOEC (Beginning of equilibrium cycle) condition, high reactivity effects of 808 ± 17 pcm and 1208 ± 17 pcm were observed for ABR-1000 MOX core and metallic core respectively. To analyze the causes of these differences in reactivity, several TRIPOLI-4 runs using mixed data libraries feature allow us to identify the nuclides and the nuclear data accounting for the major part of the observed reactivity discrepancies.

  13. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

    2002-01-01

    The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

  14. Attachment system for helium-cooled blanket of RF DEMO fusion reactor

    Energy Technology Data Exchange (ETDEWEB)

    Leshukov, A. E-mail: leshu@entek.ru; Blinov, Y.; Kovalenko, V.; Shatalov, G.; Strebkov, Y.; Strizhov, A

    2002-11-01

    The development of DEMO thermonuclear reactor is a part of Russian national program on the fusion process mastering. The DEMO-S (stationary thermonuclear reactor) should be the logic continuation of the ITER-type projects (pulse thermonuclear reactors) and the prototype for commercial power plants. DEMO reactor layout suggests to use the segmented blanket with mounting/dismounting procedure through the vacuum vessel vertical ports. Taking into account this layout the blanket attachment system has been developed and the present paper is devoted to this subject. The considered attachment system includes the lower and upper toroidal support assemblies which connect all the blanket segments in the enclosed ring. In it's turn the lower support assemblies attached to the vacuum vessel through the system of hinged support pillars. The heights of support pillars for inboard and outboard blankets are selected so that to indemnify the blanket massif thermal expansions in vertical and radial directions. The support pillars have been calculated on strength taking into account the electromagnetic loads from the plasma disruptions and blanket mass. The selection of high-strength chromium steel as a structural material for the support pillars could be considered as the results of strength analysis. The conclusions on the possibility to apply this attachment system for fusion reactor blanket and the critical issues are contained in this paper too.

  15. Gas-Cooled Thermal Reactor Program. Semiannual technical progress report, April 1, 1983-September 30, 1983

    Energy Technology Data Exchange (ETDEWEB)

    1983-12-01

    An assessment of the HTGR opportunities from the year 2000 through 2045 was the principal activity on the Market Definition Task (WBS 03). Within the Plant Technology (WBS 13) task, there were activities to develop analytical methods for investigation of Coolant Transport Behavior and to define methods and criteria for High Temperature Structural Engineering design. The activities in support of the HTGR-SC/C Lead Plant (WBS 30 and 31) were the participation in the Lead Plant System Engineering (LPSE) effort and the plant simulation task. The efforts on the Advanced HTGR systems was performed under the Modular Reactor Systems (MRS) (WBS 41) to study the potential for multiple small reactors to provide lower costs, improved safety, and higher availability than the large monolithic core reactors.

  16. Open absorption system for cooling and air conditioning using membrane contactors - Final report

    Energy Technology Data Exchange (ETDEWEB)

    Conde-Petit, M. [M. Conde Engineering, Zuerich (Switzerland); Weber, R.; Dorer, V. [Swiss Federal Laboratories for Materials Testing and Research (EMPA), Duebendorf (Switzerland)

    2008-07-01

    Air conditioning systems based upon the open absorption principle, essentially an absorption device operating at atmospheric pressure, have been proposed and investigated at many instances in the past eighty years. Their potential for improving energy efficiency is clearly recognized in the earliest research reports. By the mid 1950ies, solar thermal energy was being applied to drive open absorption-based air conditioning systems. For several reasons, however, the open absorption technology was not mature enough to take place in the mainstream. In the past two decades, vigorous efforts have been undertaken to reverse this situation, but success continued to elude, despite the fact that the main problems, such as corrosion, aerosols in the supply air, etc., have been identified. This report details the work and the main results from the MemProDEC Project. In this project innovative solutions were proposed, and successfully investigated, for the corrosion problem and the improvement of efficiency of the absorption process, in particular a new method to cool a very compact absorber. The practically uniform flow distribution for all three streams in the absorber (air, water and desiccant) warrants the contact of the air to be dehumidified with the desiccant over the whole surface of exchange (across a porous membrane). This, together with the cooling with water in counter flow to the air, are the key factors for the excellent effectiveness of the absorber. As the results show, the dehydration effectiveness of the prototype absorber is up to 150 % higher than that previously obtained by others. The solutions developed for compactness and modularity represent an important step in the way to flexible manufacturing, i.e. using a single element size to assemble autonomous air handling units of various nominal capacities. And although the manufacturing methods of the individual elements require improvement, namely by avoiding adhesive bonding, the choice of materials and the

  17. Complete Sensitivity/Uncertainty Analysis of LR-0 Reactor Experiments with MSRE FLiBe Salt and Perform Comparison with Molten Salt Cooled and Molten Salt Fueled Reactor Models

    Energy Technology Data Exchange (ETDEWEB)

    Brown, Nicholas R [ORNL; Powers, Jeffrey J [ORNL; Mueller, Don [ORNL; Patton, Bruce W [ORNL

    2016-12-01

    In September 2016, reactor physics measurements were conducted at Research Centre Rez (RC Rez) using the FLiBe (2 7LiF + BeF2) salt from the Molten Salt Reactor Experiment (MSRE) in the LR-0 low power nuclear reactor. These experiments were intended to inform on neutron spectral effects and nuclear data uncertainties for advanced reactor systems using FLiBe salt in a thermal neutron energy spectrum. Oak Ridge National Laboratory (ORNL), in collaboration with RC Rez, performed sensitivity/uncertainty (S/U) analyses of these experiments as part of the ongoing collaboration between the United States and the Czech Republic on civilian nuclear energy research and development. The objectives of these analyses were (1) to identify potential sources of bias in fluoride salt-cooled and salt-fueled reactor simulations resulting from cross section uncertainties, and (2) to produce the sensitivity of neutron multiplication to cross section data on an energy-dependent basis for specific nuclides. This report provides a final report on the S/U analyses of critical experiments at the LR-0 Reactor relevant to fluoride salt-cooled high temperature reactor (FHR) and liquid-fueled molten salt reactor (MSR) concepts. In the future, these S/U analyses could be used to inform the design of additional FLiBe-based experiments using the salt from MSRE.

  18. Dynamic Models of Heating and Cooling Coils with One—Dimensional Air Distribution

    Institute of Scientific and Technical Information of China (English)

    WangZijie; G.Krauss

    1993-01-01

    This paper presents the simulation models of the plate-fin,air-to-water(or water vapour) heat exchangers used as air-heating or air-cooling and dehumidifying colis in the HVAC(Heating,Ventilation and AIr-Conditioning)systems.The thermal models are used to calculate the heat exchange between distributing air and coil pipes and outlet temperatures of air and heat or chilled fluid.The aerodynamic models are used to account for the pressure drop of the air crossing the coil tubes,They can also be used to optimize the structres of such coils.The models are based on principal laws of teat and mass conservation and fluid mechanics.They are transparent and easy to use.In our work,a coil is considered as an assembly of numbers of basic elements in which all the state variables are unique.Therefore we can conveniently simulate the coils with different structures and different geometric parameters.Two modular programs TRNSYS(Transient System Simulation)and ESACAP are utilized as supporting softwares which make the programming and simulation greatly simplified.The coil elements and a real coil were simulated.The results were compared with the data offered by the manufacturer(company SOFICA) and also with those obtained using critical methods such as NTU method ,etc.and good agreement is attained.

  19. Study of hydrogen generation plant coupled to high temperature gas cooled reactor

    Science.gov (United States)

    Brown, Nicholas Robert

    Hydrogen generation using a high temperature nuclear reactor as a thermal driving vector is a promising future option for energy carrier production. In this scheme, the heat from the nuclear reactor drives an endothermic water-splitting plant, via coupling, through an intermediate heat exchanger. While both high temperature nuclear reactors and hydrogen generation plants have high individual degrees of development, study of the coupled plant is lacking. Particularly absent are considerations of the transient behavior of the coupled plant, as well as studies of the safety of the overall plant. The aim of this document is to contribute knowledge to the effort of nuclear hydrogen generation. In particular, this study regards identification of safety issues in the coupled plant and the transient modeling of some leading candidates for implementation in the Nuclear Hydrogen Initiative (NHI). The Sulfur Iodine (SI) and Hybrid Sulfur (HyS) cycles are considered as candidate hydrogen generation schemes. Several thermodynamically derived chemical reaction chamber models are coupled to a well-known reference design of a high temperature nuclear reactor. These chemical reaction chamber models have several dimensions of validation, including detailed steady state flowsheets, integrated loop test data, and bench scale chemical kinetics. Eight unique case studies are performed based on a thorough literature review of possible events. The case studies are: (1) feed flow failure from one section of the chemical plant to another, (2) product flow failure (recycle) within the chemical plant, (3) rupture or explosion within the chemical plant, (4) nuclear reactor helium inlet overcooling due to a process holding tank failure, (5) helium inlet overcooling as an anticipated transient without SCRAM, (6) total failure of the chemical plant, (7) parametric study of the temperature in an individual reaction chamber, and (8) control rod insertion in the nuclear reactor. Various parametric

  20. Production of liquid fuels with a high-temperature gas-cooled reactor

    Science.gov (United States)

    Quade, R. N.; Vrable, D. L.; Green, L., Jr.

    An exploration is made of the technical, economic and environmental impact feasibility of integrating coal liquefaction methods directly and indirectly with a nuclear reactor source of process heat, with stress on the production of synthetic jet fuel. Production figures and operating costs are compared for indirect conventional and nuclear processes using Lurgi-Fischer-Tropsch technology with direct conventional and nuclear techniques employing the advanced SRC-II technology, and it is concluded that significant advantages in coal savings and environmental impact can be expected from nuclear reactor integration.

  1. Pilot-plant testing of magnetic filters for the N-Reactor primary cooling circuit

    Energy Technology Data Exchange (ETDEWEB)

    Emory, B.B.

    1982-01-01

    Data obtained during the laboratory loop test program using the high power HGMF indicates that removal efficiency for /sup 60/Co and subsequently the bulk of the crud, will be greater than 90% at field strength above .1 Tesla for the expanded metal mesh matrix. However, since /sup 54/Mn seems to exhibit paramagnetic behavior and the possibility of quantities of alpha iron forming during reactor shut down from oxygen inleakage, a field strength of .5 to 1 Tesla may be more appropriate for a full scale on-reactor installation. Crud loading of 50 gm per kg of matrix weight are readily obtainable and up to twice that amount has been reached.

  2. A simplified Probabilistic Safety Assesment of a Steam-Methane Reforming Hydrogen Production Plant coupled to a High-Temperature Gas Cooled Nuclear Reactor

    OpenAIRE

    Nelson Edelstein, Pamela; Flores Flores, Alain; Francois Lacouture, Juan Luis

    2005-01-01

    A Probabilistic Safety Assessment (PSA) is being developed for a steam-methane reforming hydrogen production plant linked to a High-Temperature Gas Cooled Nuclear Reactor (HTGR). This work is based on the Japan Atomic Energy Research Institute’s (JAERI) High Temperature Test Reactor (HTTR) prototype in Japan. This study has two major objectives: calculate the risk to onsite and offsite individuals, and calculate the frequency of different types of damage to the complex. A simplified HAZOP...

  3. The European JASMIN Project for the Development of a New Safety Simulation Code, ASTEC-Na, for Na-cooled Fast Neutron Reactors

    OpenAIRE

    GIRAULT N.; VAN DORSSELAERE J.p.; Jacq, F.; BRILLANT G.; KISSANE Martin; BANDINI, G; Buck,M.; CHAMPIGNY J.; Hering, W; Perez-Martin, S.; Herranz, L; RAISON Philippe; Reinke, N; TUCEK Kamil; VERWAERDE D.

    2012-01-01

    The 4-year JASMIN collaborative project, involving 9 organizations, was launched by IRSN end of 2011 within the 7th European R&D Framework Programme on the enhancement of Na-cooled Fast Neutron Reactors (SFR) safety for a higher resistance to severe accidents. The project aims at developing a new European simulation code, ASTEC-Na, with a modern architecture, sufficiently flexible to account for innovative reactor designs and eventually new types of fuel and claddings and accounting for resul...

  4. Analysis of an air cooled ammonia-water vertical tubular absorber

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez-Seara, J.; Uhia, F.J.; Sieres, J. [Area de Maquinas y Motores Termicos, Escuela Tecnica Superior de Ingenieros Industriales, Campus Lagoas-Marcosende, No 9, 36310 Vigo (Spain)

    2007-03-15

    This paper presents a detailed analysis of an ammonia-water vertical tubular absorber cooled by air. The absorption process takes place co-currently upward inside the tubes. The tubes are externally finned with continuous plate fins and the tube rows are arranged staggered in the direction of the air flow. The air is forced over the tube bank and circulates between the plain fins in cross flow with the ammonia-water mixture. The analysis has been carried out by means of a mathematical model developed on the basis of mass and energy balances and heat and mass transfer equations. The model takes into account separately the churn, slug and bubbly flow patterns experimentally forecasted in this type of absorption processes inside vertical tubes and considers the simultaneous heat and mass transfer processes in both liquid and vapour phases, as well as heat transfer to the cooling air. The model has been implemented in a computer program. Results based on a representative design and nominal operating conditions of an absorber for a small capacity ammonia-water absorption refrigeration system are shown. A parametric analysis was realised to investigate the influence of the design parameters and operating conditions on the absorber performance. The noteworthy results that have effect on practical design of the absorber are presented and commented. (authors)

  5. Saturated Adaptive Output-Feedback Power-Level Control for Modular High Temperature Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2014-11-01

    Full Text Available Small modular reactors (SMRs are those nuclear fission reactors with electrical output powers of less than 300 MWe. Due to its inherent safety features, the modular high temperature gas-cooled reactor (MHTGR has been seen as one of the best candidates for building SMR-based nuclear plants with high safety-level and economical competitive power. Power-level control is crucial in providing grid-appropriation for all types of SMRs. Usually, there exists nonlinearity, parameter uncertainty and control input saturation in the SMR-based plant dynamics. Motivated by this, a novel saturated adaptive output-feedback power-level control of the MHTGR is proposed in this paper. This newly-built control law has the virtues of having relatively neat form, of being strong adaptive to parameter uncertainty and of being able to compensate control input saturation, which are given by constructing Lyapunov functions based upon the shifted-ectropies of neutron kinetics and reactor thermal-hydraulics, giving an online tuning algorithm for the controller parameters and proposing a control input saturation compensator respectively. It is proved theoretically that input-to-state stability (ISS can be guaranteed for the corresponding closed-loop system. In order to verify the theoretical results, this new control strategy is then applied to the large-range power maneuvering control for the MHTGR of the HTR-PM plant. Numerical simulation results show not only the relationship between regulating performance and control input saturation bound but also the feasibility of applying this saturated adaptive control law practically.

  6. Safety properties of sodium-cooled fast reactors%钠冷快堆及其安全特性

    Institute of Scientific and Technical Information of China (English)

    徐銤; 杨红义

    2016-01-01

    钠冷快堆是第四代核能系统国际论坛(GIF)公布的6种第四代先进反应堆中研发进展最快、最接近满足商业核电厂需要的堆型。钠冷快堆因其在固有安全性以及可增殖核燃料、嬗变长寿命放射性废物等方面的优势,得到了世界各国的重视。文章以中国第一座钠冷快堆——中国实验快堆(China Experimental Fast Reactor,CEFR)为例,介绍了钠冷快堆在设计及运行方面的安全特性。%The sodium-cooled fast reactor is the fastest prototype and the closest to com-mercialization for nuclear power plants amongst the six types of fourth generation reactors, as an-nounced at the Generation IV International Forum. Many countries are paying more and more at-tention to the research and development of these reactors, due to the inherent safety features, effi-cient utilization of uranium with the breeding of the plutonium, and transmutation of long-lived ac-tinides. The design and operational safety characteristics of the China Experimental Fast Reactor are reviewed in this paper.

  7. Operation of Fusion Reactors in One Atmosphere of Air Instead of Vacuum Systems

    Science.gov (United States)

    Roth, J. Reece

    2009-07-01

    Engineering design studies of both magnetic and inertial fusion power plants have assumed that the plasma will undergo fusion reactions in a vacuum environment. Operation under vacuum requires an expensive additional major system for the reactor-a vacuum vessel with vacuum pumping, and raises the possibility of sudden unplanned outages if the vacuum containment is breached. It would be desirable in many respects if fusion reactors could be made to operate at one atmosphere with air surrounding the plasma, thus eliminating the requirement of a pressure vessel and vacuum pumping. This would have obvious economic, reliability, and engineering advantages for currently envisaged power plant reactors; it would make possible forms of reactor control not possible under vacuum conditions (i.e. adiabatic compression of the fusion plasma by increasing the pressure of surrounding gas); it would allow reactors used as aircraft engines to operate as turbojets or ramjets in the atmosphere, and it would allow reactors used as fusion rockets to take off from the surface of the earth instead of low earth orbit.

  8. A preliminary approach to the extension of the Transuranus code to the fuel rod performance analysis of HLM-cooled nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Luzzi, L.; Botazzoli, P.; Devita, M.; Di Marcello, V.; Pastore, G. [Department of Energy, Politecnico di Milano, Enrico Fermi Center for Nuclear Studies - CeSNEF, via Ponzio 34/3, 20133 Milano (Italy)

    2010-07-01

    This paper briefly presents a preliminary modelling approach, aimed at the extension of the TRANSURANUS code to the fuel rod performance analysis of Heavy Liquid Metal (HLM) cooled nuclear reactors, with specific reference to the employment of the T91 steel as cladding material and of the liquid Lead-Bismuth Eutectic (LBE) as coolant. On the basis of literature indications, correlations for heat transfer to LBE, corrosion behaviour and thermo-mechanical properties of T91 are proposed, and some open issues are discussed in prospect of more reliable fuel rod performance analysis of HLM-cooled nuclear reactors. (authors)

  9. Effect of Chromium on CCT Diagrams of Novel Air-Cooled Bainite Steels Analyzed by Neural Network

    Institute of Scientific and Technical Information of China (English)

    YOU Wei; XU Wei-hong; LIU Ya-xiu; BAI Bing-zhe; FANG Hong-sheng

    2007-01-01

    The quantitative effects of chromium content on continuous cooling transformation (CCT) diagrams of novel air-cooled bainite steels were analyzed using artificial neural network models. The results showed that the chromium may retard the high and medium-temperature martensite transformation.

  10. Literature review on metallic fuel source term for sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Nam Duk; Bae, Moo Hoon; Shin, An Dong; Huh, Chang Wook [Korea Institute of Nuclear Safety, Daejon (Korea, Republic of)

    2012-10-15

    Source term is defined as the release of radionuclides from the fuel and coolant into the containment, and subsequently to the environment, following a severe accident where a significant portion of the reactor core has melted. Of the many issues associated with the development and deployment of SFRs, one of high regulatory importance is the source term to be used in the siting of the reactor. Apart from assessing the radiological consequences for siting, it is also important for designing filtering systems and even reactor components. Overly conservative source term for light water reactor, TID 14844 demands for very fast closure of main steam isolation valves, rapid startup of emergency diesels, and safety systems designed to mitigate gaseous iodine. In spite of this importance, most of the knowledge we have for SFR source term comes from the research performed before 1980s. Moreover, majority of the work on metallic fuels was done during the late 1950's through the 1960's. This paper reviews and summarizes the main characteristics of SFR source terms based on the available literatures.

  11. Performance of low smeared density sodium-cooled fast reactor metal fuel

    Science.gov (United States)

    Porter, D. L.; Chichester, H. J. M.; Medvedev, P. G.; Hayes, S. L.; Teague, M. C.

    2015-10-01

    An experiment was performed in the Experimental Breeder Rector-II (EBR-II) in the 1990s to show that metallic fast reactor fuel could be used in reactors with a single, once-through core. To prove the long duration, high burnup, high neutron exposure capability an experiment where the fuel pin was designed with a very large fission gas plenum and very low fuel smeared density (SD). The experiment, X496, operated to only 8.3 at.% burnup because the EBR-II reactor was scheduled for shut-down at that time. Many of the examinations of the fuel pins only funded recently with the resurgence of reactor designs using very high-burnup fuel. The results showed that, despite the low smeared density of 59% the fuel swelled radially to contact the cladding, fission gas release appeared to be slightly higher than demonstrated in conventional 75%SD fuel tests and axial growth was about the same as 75% SD fuel. There were axial positions in some of the fuel pins which showed evidence of fuel restructuring and an absence of fission products with low melting points and gaseous precursors (Cs and Rb). A model to investigate whether these areas may have overheated due to a loss of bond sodium indicates that it is a possible explanation for the fuel restructuring and something to be considered for fuel performance modeling of low SD fuel.

  12. High Temperature Fusion Reactor Cooling Using Brayton Cycle Based Partial Energy Conversion

    Science.gov (United States)

    Juhasz, Albert J.; Sawicki, Jerzy T.

    2003-01-01

    For some future space power systems using high temperature nuclear heat sources most of the output energy will be used in other than electrical form, and only a fraction of the total thermal energy generated will need to be converted to electrical work. The paper describes the conceptual design of such a partial energy conversion system, consisting of a high temperature fusion reactor operating in series with a high temperature radiator and in parallel with dual closed cycle gas turbine (CCGT) power systems, also referred to as closed Brayton cycle (CBC) systems, which are supplied with a fraction of the reactor thermal energy for conversion to electric power. Most of the fusion reactor's output is in the form of charged plasma which is expanded through a magnetic nozzle of the interplanetary propulsion system. Reactor heat energy is ducted to the high temperature series radiator utilizing the electric power generated to drive a helium gas circulation fan. In addition to discussing the thermodynamic aspects of the system design the authors include a brief overview of the gas turbine and fan rotor-dynamics and proposed bearing support technology along with performance characteristics of the three phase AC electric power generator and fan drive motor.

  13. Optimization of Overall Project of ChinaLEad-Alloy Cooled Reactor

    Institute of Scientific and Technical Information of China (English)

    DUAN; Tian-ying; LONG; Bin; HU; Yun; LIU; Xing-ming; SUN; Gang; LIU; Yi-zhe; YANG; Yong; PEI; Zhi-yong; FENG; Wei-wei; YU; Xiao

    2012-01-01

    <正>The application of accelerator driven sub-critical reactor which is able to transmute spent fuel and long-lived radioactive fission products is highly concerned in the international research projects. Lots of researches have been done in many countries. While China Institute of Atomic Energy has built a fast

  14. Sustainability and Efficiency Improvements of Gas-Cooled High Temperature Reactors

    NARCIS (Netherlands)

    Marmier, A.

    2012-01-01

    The work presented in this thesis covers three fundamental aspects of High Temperature Reactor (HTR) performance, namely fuel testing under irradiation for maximized safety and sustainability, fuel architecture for improved economy and sustainability, and a novel Balance of Plant concept to enable f

  15. Porosity Effect in the Core Thermal Hydraulics for Ultra High Temperature Gas-cooled Reactor

    Directory of Open Access Journals (Sweden)

    Motoo Fumizawa

    2008-12-01

    Full Text Available This study presents an experimental method of porosity evaluation and a predictive thermal-hydraulic analysis with packed spheres in a nuclear reactor core. The porosity experiments were carried out in both a fully shaken state with the closest possible packing and in a state of non-vibration. The predictive analysis considering the fixed porosity value was applied as a design condition for an Ultra High Temperature Reactor Experiment (UHTREX. The thermal-hydraulic computer code was developed and identified as PEBTEMP. The highest outlet coolant temperature of 1316 oC was achieved in the case of an UHTREX at Los Alamos Scientific Laboratory, which was a small scale UHTR. In the present study, the fuel was changed to a pebble type, a porous media. In order to compare the present pebble bed reactor and UHTREX, a calculation based on HTGR-GT300 was carried out in similar conditions with UHTREX; in other words, with an inlet coolant temperature of 871oC, system pressure of 3.45 MPa and power density of 1.3 w/cm3. As a result, the fuel temperature in the present pebble bed reactor showed an extremely lower value compared to that of UHTREX.

  16. Ferritic steels for sodium-cooled fast reactors: Design principles and challenges

    Science.gov (United States)

    Raj, Baldev; Vijayalakshmi, M.

    2010-09-01

    An overview of the current status of development of ferritic steels for emerging fast reactor technologies is presented in this paper. The creep-resistant 9-12Cr ferritic/martensitic steels are classically known for steam generator applications. The excellent void swelling resistance of ferritic steels enabled the identification of their potential for core component applications of fast reactors. Since then, an extensive knowledge base has been generated by identifying the empirical correlations between chemistry of the steels, heat treatment, structure, and properties, in addition to their in-reactor behavior. A few concerns have also been identified which pertain to high-temperature irradiation creep, embrittlement, Type IV cracking in creep-loaded weldments, and hard zone formation in dissimilar joints. The origin of these problems and the methodologies to overcome the limitations are highlighted. Finally, the suitability of the ferritic steels is re-evaluated in the emerging scenario of the fast reactor technology, with a target of achieving better breeding ratio and improved thermal efficiency.

  17. Constructal design of finned tubes used in air-cooled heat exchangers

    Energy Technology Data Exchange (ETDEWEB)

    Shokouhmand, Hossein; Mahjoub, Shoeib [University of Tehran, Tehran (Iran, Islamic Republic of); Salimpour, Mohammad Reza [Isfahan University of Technology, Isfahan (Iran, Islamic Republic of)

    2014-06-15

    The present study documents the constructal design and optimization of finned tubes used in air-cooled heat exchangers. The considered tubes are equipped with annular fins. The aim is to minimize the overall thermal resistance by morphing the geometry. The geometrical and thermo-physical parameters considered are the number of fins, ratio of fin height to tube diameter, Stanton number, ratio of fin conductivity to air conductivity, ratio of in-tube fluid conductivity to air conductivity and dimensionless pressure drop. Two constraints are applied in the optimization process: fixed overall volume of heat exchanger and fixed volume fraction of fin material. It is found that there exist optimal values for the number and the height of fins. Moreover, the optimal heat transfer has an extremum in a special volume fraction of fin material.

  18. Effects of Using Wind Guiding Nets to Improve Prevailing Ambient Wind on Direct Air-cooled Condensers

    Institute of Scientific and Technical Information of China (English)

    YANG Jianguo; LIU Da; ZHANG Zhaoying; JU Xiaoli

    2012-01-01

    Prevailing ambient wind is the main reason that causes inlet flow rate (air mass flow rate) decreasing and air flowing backward to the air-cooled condenser fans upward to the wind, hence a set of wind guiding nets is designed to improve the detrimental effect. Fig. 1 shows four typical units of a 1000MW direct air-cooled condenser (DACC) and a set of wind guiding nets installed under its edge upward to the ambient wind. As shown in Fig. 2, the fan inlet flow rate decreases as the prevailing ambient wind velocity increasing, especially for the first two units upward to the wind.

  19. CFD simulations of moderator flow inside Calandria of the Passive Moderator Cooling System of an advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pal, Eshita [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Kumar, Mukesh [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Joshi, Jyeshtharaj B., E-mail: jbjoshi@gmail.com [Homi Bhabha National Institute, Anushaktinagar, Mumbai 400 094 (India); Department of Chemical Engineering, Institute of Chemical Technology, Matunga, Mumbai 400019 India (India); Nayak, Arun K. [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India); Vijayan, Pallippattu K., E-mail: vijayanp@barc.gov.in [Reactor Engineering Division, Bhabha Atomic Research Center, Trombay, Mumbai 400 085 (India)

    2015-10-15

    Highlights: • CFD simulations in the Calandria of an advanced reactor under natural circulation. • Under natural convection, majority of the flow recirculates within the Calandria. • Maximum temperature is located at the top and center of the fuel channel matrix. • During SBO, temperature inside Calandria is stratified. - Abstract: Passive systems are being examined for the future Advanced Nuclear Reactor designs. One of such concepts is the Passive Moderator Cooling System (PMCS), which is designed to remove heat from the moderator in the Calandria vessel passively in case of an extended Station Black Out condition. The heated heavy-water moderator (due to heat transferred from the Main Heat Transport System (MHTS) and thermalization of neutrons and gamma from radioactive decay of fuel) rises upward due to buoyancy, gets cooled down in a heat exchanger and returns back to Calandria, completing a natural circulation loop. The natural circulation should provide sufficient cooling to prevent the increase of moderator temperature and pressure beyond safe limits. In an earlier study, a full-scale 1D transient simulation was performed for the reactor including the MHTS and the PMCS, in the event of a station blackout scenario (Kumar et al., 2013). The results indicate that the systems remain within the safe limits for 7 days. However, the flow inside a geometry like Calandria is quite complex due to its large size and inner complexities of dense fuel channel matrix, which was simplified as a 1D pipe flow in the aforesaid analysis. In the current work, CFD simulations are performed to study the temperature distributions and flow distribution of moderator inside the Calandria vessel using a three-dimensional CFD code, OpenFoam 2.2.0. First, a set of steady state simulation was carried out for a band of inlet mass flow rates, which gives the minimum mass flow rate required for removing the maximum heat load, by virtue of prediction of hot spots inside the Calandria

  20. Methodology for uncertainty calculation of net total cooling effect estimation for rating room air conditioners and packaged terminal air conditioners

    Energy Technology Data Exchange (ETDEWEB)

    Fonseca Diaz, Nestor [Universidad Tecnologica de Pereira, Facultad de Ingenieria Mecanica, Pereira (Colombia); University of Liege, Campus du Sart Tilman, Bat: B49, P33, B-4000 Liege (Belgium)

    2009-09-15

    This article presents the general procedure for uncertainty calculation of net total cooling effect estimation for rating room air conditioners and packaged terminal air conditioners, by means of measurements carried out in a test bench specially designed for this purpose. The uncertainty analysis presented in this work looks for establishing a confidence degree or certainty of experimental results. It is particularly important considering that international standards related to this type of analysis are too ambiguous when treating this subject. The uncertainty analysis is on the other hand an indispensable requirement to international standard ISO 17025 [ISO, 2005. International Standard. 17025. General Requirement to Test and Calibration Laboratories Competences. International Organization for Standardization, Geneva.], which must be applied to obtain the required quality levels according to the Word Trade Organization WTO. (author)

  1. Computational Fluid Dynamics Analyses on Very High Temperature Reactor Air Ingress

    Energy Technology Data Exchange (ETDEWEB)

    Chang H Oh; Eung S. Kim; Richard Schultz; David Petti; Hyung S. Kang

    2009-07-01

    A preliminary computational fluid dynamics (CFD) analysis was performed to understand density-gradient-induced stratified flow in a Very High Temperature Reactor (VHTR) air-ingress accident. Various parameters were taken into consideration, including turbulence model, core temperature, initial air mole-fraction, and flow resistance in the core. The gas turbine modular helium reactor (GT-MHR) 600 MWt was selected as the reference reactor and it was simplified to be 2-D geometry in modeling. The core and the lower plenum were assumed to be porous bodies. Following the preliminary CFD results, the analysis of the air-ingress accident has been performed by two different codes: GAMMA code (system analysis code, Oh et al. 2006) and FLUENT CFD code (Fluent 2007). Eventually, the analysis results showed that the actual onset time of natural convection (~160 sec) would be significantly earlier than the previous predictions (~150 hours) calculated based on the molecular diffusion air-ingress mechanism. This leads to the conclusion that the consequences of this accident will be much more serious than previously expected.

  2. Using cathode spacers to minimize reactor size in air cathode microbial fuel cells

    KAUST Repository

    Yang, Qiao

    2012-04-01

    Scaling up microbial fuel cells (MFCs) will require more compact reactor designs. Spacers can be used to minimize the reactor size without adversely affecting performance. A single 1.5mm expanded plastic spacer (S1.5) produced a maximum power density (973±26mWm -2) that was similar to that of an MFC with the cathode exposed directly to air (no spacer). However, a very thin spacer (1.3mm) reduced power by 33%. Completely covering the air cathode with a solid plate did not eliminate power generation, indicating oxygen leakage into the reactor. The S1.5 spacer slightly increased columbic efficiencies (from 20% to 24%) as a result of reduced oxygen transfer into the system. Based on operating conditions (1000ς, CE=20%), it was estimated that 0.9Lh -1 of air would be needed for 1m 2 of cathode area suggesting active air flow may be needed for larger scale MFCs. © 2012 Elsevier Ltd.

  3. Transfair. An air method of floor heating and cooling; Transfair. Procede de chauffage et de rafraichissement par le sol a air

    Energy Technology Data Exchange (ETDEWEB)

    Desvouas, C. [Tarnsfair TTR, 77 - Perthes en Gatinais (France)

    1997-12-31

    This paper presents the `Transfair` method which consists in the use of air instead of water in heating and cooling systems. This presentation comprises 4 parts. Part 1 is a general presentation of the activities of the French TTR company which has developed the Transfair method and its realizations in industrial space heating and in chemical industry (refrigeration units for sulfuric acid). Part 2 is a comparative evaluation of equivalent water and air systems in order to emphasize the advantages of air cooling systems (simplicity and reliability). Part 3 is a presentation of a software for the optimization of the dimensioning of components and investment costs of industrial air space heating systems. Part 4 is a presentation of the feasibility study of a floor cooled by air circulation and with a self-balancing of circuits confirmed by flow rate measurements. (J.S.)

  4. Comfort air temperature influence on heating and cooling loads of a residential building

    Science.gov (United States)

    Stanciu, C.; Șoriga, I.; Gheorghian, A. T.; Stanciu, D.

    2016-08-01

    The paper presents the thermal behavior and energy loads of a two-level residential building designed for a family of four, two adults and two students, for different inside comfort levels reflected by the interior air temperature. Results are intended to emphasize the different thermal behavior of building elements and their contribution to the building's external load. The most important contributors to the building thermal loss are determined. Daily heating and cooling loads are computed for 12 months simulation in Bucharest (44.25°N latitude) in clear sky conditions. The most important aspects regarding sizing of thermal energy systems are emphasized, such as the reference months for maximum cooling and heating loads and these loads’ values. Annual maximum loads are encountered in February and August, respectively, so these months should be taken as reference for sizing thermal building systems, in Bucharest, under clear sky conditions.

  5. Annular core for Modular High-Temperature Gas-Cooled Reactor (MHTGR)

    Energy Technology Data Exchange (ETDEWEB)

    Turner, R.F.; Baxter, A.M.; Stansfield, O.M.; Vollman, R.E.

    1987-08-01

    The active core of the 350 MW(t) MHTGR is annular in configuration, shaped to provide a large external surface-to-volume ratio for the transport of heat radially to the reactor vessel in case of a loss of coolant flow. For a given fuel temperature limit, the annular core provides approximately 40% greater power output over a typical cylindrical configuration. The reactor core is made up of columns of hexagonal blocks, each 793-mm high and 360-mm wide. The active core is 3.5 m in o.d., 1.65 m in i.d., and 7.93-m tall. Fuel elements contain TRISO-coated microspheres of 19.8% enriched uranium oxycarbide and of fertile thorium oxide. The core is controlled by 30 control rods which enter the inner and outer side reflectors from above.

  6. Annular core for the Modular High-Temperature Gas-cooled Reactor (MHTGR)

    Energy Technology Data Exchange (ETDEWEB)

    Turner, R.F.; Baxter, A.M.; Stansfield, O.M.; Vollman, R.E.

    The active core of the 350 MW(t) MHTGR is annular in configuration, shaped to provide a large external surface-to-volume ratio for the transport of heat radially to the reactor vessel in case of a loss of coolant flow. For a given fuel temperature limit, the annular core provides approximately 40% greater power output over a typical cylindrical configuration. The reactor core is made up to columns of hexagonal blocks, each 793 mm high and 360 mm wide. The active core is 3.5 m in outside diameter, 1.65 m in inside diameter, and 7.93 m tall. Fuel elements contain TRISO-coated microspheres of 19.8% enriched uranium oxycarbide and of fertile thorium oxide. The core is controlled by 30 control rods which enter the inner and outer side reflectors from above.

  7. Design of a Low Power, Fast-Spectrum, Liquid-Metal Cooled Surface Reactor System

    Science.gov (United States)

    Marcille, T. F.; Dixon, D. D.; Fischer, G. A.; Doherty, S. P.; Poston, D. I.; Kapernick, R. J.

    2006-01-01

    In the current 2005 US budget environment, competition for fiscal resources make funding for comprehensive space reactor development programs difficult to justify and accommodate. Simultaneously, the need to develop these systems to provide planetary and deep space-enabling power systems is increasing. Given that environment, designs intended to satisfy reasonable near-term surface missions, using affordable technology-ready materials and processes warrant serious consideration. An initial lunar application design incorporating a stainless structure, 880 K pumped NaK coolant system and a stainless/UO2 fuel system can be designed, fabricated and tested for a fraction of the cost of recent high-profile reactor programs (JIMO, SP-100). Along with the cost reductions associated with the use of qualified materials and processes, this design offers a low-risk, high-reliability implementation associated with mission specific low temperature, low burnup, five year operating lifetime requirements.

  8. Review of core disruptive accident analysis for liquid-metal cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. C.; Na, B. C.; Hahn, D. H

    1997-04-01

    Analysis methodologies of core disruptive accidents (CDAs) are reviewed. The role of CDAS in the overall safety evaluation of fast reactors has not always been well defined nor universally agreed upon. However, they have become a traditional issue in LMR safety, design, and licensing. The study is for the understanding of fast reactor behavior under CDA conditions to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting design features for the KALIMER developments. The methods used to analyze CDAs from initiating event to complete core disruption are described. Two examples of CDA analyses for CRBRP and ALMR are given and R and D needed for better understanding of CDA phenomena are proposed. (author). 10 refs., 2 tabs., 3 figs

  9. Air trichloroethylene oxidation in a corona plasma-catalytic reactor

    Science.gov (United States)

    Masoomi-Godarzi, S.; Ranji-Burachaloo, H.; Khodadadi, A. A.; Vesali-Naseh, M.; Mortazavi, Y.

    2014-08-01

    The oxidative decomposition of trichloroethylene (TCE; 300 ppm) by non-thermal corona plasma was investigated in dry air at atmospheric pressure and room temperature, both in the absence and presence of catalysts including MnOx, CoOx. The catalysts were synthesized by a co-precipitation method. The morphology and structure of the catalysts were characterized by BET surface area measurement and Fourier Transform Infrared (FTIR) methods. Decomposition of TCE and distribution of products were evaluated by a gas chromatograph (GC) and an FTIR. In the absence of the catalyst, TCE removal is increased with increases in the applied voltage and current intensity. Higher TCE removal and CO2 selectivity is observed in presence of the corona and catalysts, as compared to those with the plasma alone. The results show that MnOx and CoOx catalysts can dissociate the in-plasma produced ozone to oxygen radicals, which enhances the TCE decomposition.

  10. Patent Analysis of Ferritic/Martensitic Steels for the Fuel Cladding in Sodium-cooled Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Jong Hyuk; Kim, Sung Ho; Kim, Tae Kyu; Kim, Woo Gon; Jang, Jin Sung; Kim, Dae Whan; Han, Chang Hee; Lee, Chan Bock

    2007-09-15

    The Korean, Japanese, U.S. and European patents related to the ferritic/martensitic steels were systematically surveyed to evaluate their patent status, which would be applicable to the fuel cladding materials for the Sodium-cooled Fast Reactor (SFR). From the surveys, totally 38 patents were finally selected for the quantitative and qualitative analysis. Among them, 28 patents (74%) were processed by Japanese companies and Sumitomo Metal industries Ltd. was top-ranked in the number (9) of priority patents. On the basis of these surveys, most patents could be applicable to the fuel cladding materials for SFR and, especially, some useful patents as the cladding were registered by the Russian and the Korean.

  11. Analysis on thermophoretic deposit of fine particle on water wall of 10 MW high temperature gas-cooled reactor

    Institute of Scientific and Technical Information of China (English)

    ZHOU Tao; YANG Rui-Chang; JIA Dou-Nan

    2005-01-01

    The water wall is an important part of the passive natural circulation residual heat removal system in a high temperature gas-cooled reactor. The maximum temperatures of the pressure shell and the water wall are calculated using annular vertical closed cavity model. Fine particles can deposit on the water wall due to the thermophore sis effect. This deposit can affect heat transfer. The thermophoretic deposit efficiency is calculated by using Batch and Shen's formula fitted for both laminar flow and turbulent flow. The calculated results indicate that natural convection is turbulent in the closed cavity. The transient thermophoretic deposit efficiency rises with the increase of the pressure shell's temperature. Its maximum value is 14%.

  12. Phase characteristics of rare earth elements in metallic fuel for a sodium-cooled fast reactor by injection casting

    Science.gov (United States)

    Kuk, Seoung Woo; Kim, Ki Hwan; Kim, Jong Hwan; Song, Hoon; Oh, Seok Jin; Park, Jeong-Yong; Lee, Chan Bock; Youn, Young-Sang; Kim, Jong-Yun

    2017-04-01

    Uranium-zirconium-rare earth (U-Zr-RE) fuel slugs for a sodium-cooled fast reactor were manufactured using a modified injection casting method, and investigated with respect to their uniformity, distribution, composition, and phase behavior according to RE content. Nd, Ce, Pr, and La were chosen as four representative lanthanide elements because they are considered to be major RE components of fuel ingots after pyroprocessing. Immiscible layers were found on the top layers of the melt-residue commensurate with higher fuel slug RE content. Scanning electron microscopy-energy-dispersive X-ray spectroscopy (SEM-EDS) data showed that RE elements in the melt-residue were distributed uniformly throughout the fuel slugs. RE element agglomeration did not contaminate the fuel slugs but strongly affected the RE content of the slugs.

  13. Analysis of two-phase flow instability in helical tube steam generator in high temperature gas cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Yu; Lv, Xuefeng; Wang, Shengfei; Niu, Fenglei; Tian, Li [North China Electric Power Univ., Beijing (Switzerland)

    2012-03-15

    The steam generator composed of multi-helical tubes is used in high temperature gas cooled reactors and two-phase flow instability should be avoided in design. And density-wave oscillation which is mainly due to flow, density and the relationship between the pressure drop delays and feedback effects is one of the two-phase flow instability phenomena easily to occur. Here drift-flux model is used to simulate the performance of the fluid in the secondary side and frequency domain and time domain methods are used to evaluate whether the density-wave oscillation will happen or not. Several operating conditions with nominal power from 15% to 30% are calculated in this paper. The results of the two methods are in accordance, flow instability will occur when power is less than 20% nominal power, which is also according with the result of the experiments well.

  14. Study of the Effect of Reduced Iron Temperature Rising on Total Carbon Formation in Iron Reactor Isobaric and Cooling Zone

    Directory of Open Access Journals (Sweden)

    Bayu Alamsari

    2010-01-01

    Full Text Available We presented the mathematical model in the iron reactor. The model was limited to Isobaric Zone and Cooling Zone termed as IZ and CZ, respectively. The simulation was done by adapting the heat and mass transfer equations. The main purpose of this paper is to estimate the temperature increasing effect of reduced-iron on sponge-iron quality. The calculations are solved using Finite Element Method (FEM. The results showed that the temperature and concentration values from the simulation have high similarity to the reference data with Root-Mean-Square Error (RMSE about 0.7. The formation of total-carbon in the both zones decreased metallization degree until 1.72%. The increase in reduced-iron temperature higher than 1200 K produces total-carbon higher than 3%. Thus the increase in reduced-iron temperature more than 1200 K is not recommended because it can decrease metallization degree.

  15. Studies on disintegrating spherical fuel elements of high temperature gas-cooled reactor by a electrochemical method

    Science.gov (United States)

    Tian, Lifang; Wen, Mingfen; Chen, Jing

    2013-01-01

    Spherical fuel elements of a high temperature gas-cooled reactor were disintegrated through a electrochemical method with NaNO3 as electrolyte. The X-ray diffraction spectra and total carbon contents of the graphite fragments were determined, and the results agreed with those from simulated fuel elements. After conducting the characterization analysis and the leaching experiment of coated fuel particles, the uranium concentrations of leaching solutions and spent electrolyte were found to be at background levels. The results demonstrate the effectiveness of the improved electrochemical method with NaNO3 as electrolyte in disintegrating the unirradiated fuel elements without any damage to the coated fuel particles. Moreover, the method avoided unexpected radioactivity contamination to the graphite matrix and spent electrolyte.

  16. Analysis of dashpot performance for rotating control drums of a lithium cooled fast reactor concept

    Science.gov (United States)

    Wenzler, C. J.

    1972-01-01

    A dashpot was incorporated in the design of the drive train of the rotating control drum to prevent shock damage to the control drum and drive train at the termination of a scram action. A rotating vane dashpot using reactor coolant lithium as a damping fluid appears to be the best candidate of the various damping devices explored. A performance analysis, results and discussion of vane type dashpots are presented.

  17. Implementation Plan for Qualification of Sodium-Cooled Fast Reactor Technology Information

    Energy Technology Data Exchange (ETDEWEB)

    Moe, Wayne [Idaho National Lab. (INL), Idaho Falls, ID (United States); Honma, George [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-03-01

    This document identifies and discusses implementation elements that can be used to facilitate consistent and systematic evaluation processes relating to quality attributes of technical information (with focus on SFR technology) that will be used to support licensing of advanced reactor designs. Information may include, but is not limited to, design documents for SFRs, research-and-development (R&D) data and associated documents, test plans and associated protocols, operations and test data, international research data, technical reports, and information associated with past U.S. Nuclear Regulatory Commission (NRC) reviews of SFR designs. The approach for determining acceptability of test data, analysis, and/or other technical information is based on guidance provided in INL/EXT-15-35805, “Guidance on Evaluating Historic Technology Information for Use in Advanced Reactor Licensing.” The implementation plan can be adopted into a working procedure at each of the national laboratories performing data qualification, or by applicants seeking future license application for advanced reactor technology.

  18. Homogenization of the internal structures of a reactor with the cooling fluid

    Energy Technology Data Exchange (ETDEWEB)

    Robbe, M.F. [CEA Saclay, SEMT, 91 - Gif sur Yvette (France); Bliard, F. [Socotec Industrie, Service AME, 78 - Montigny le Bretonneux (France)

    2001-07-01

    To take into account the influence of a structure net among a fluid flow, without modelling exactly the structure shape, a concept of ''equivalent porosity method'' was developed. The structures are considered as solid pores inside the fluid. The structure presence is represented by three parameters: a porosity, a shape coefficient and a pressure loss coefficient. The method was studied for an Hypothetical Core Disruptive Accident in a Liquid Metal Fast Breeder Reactor, but it can be applied to any problem involving fluid flow getting through a solid net. The model was implemented in the computer code CASTEM-PLEXUS and validated on an analytical shock tube test, simulating an horizontal slice of a schematic LMFBR in case of a HCDA (bubble at high pressure, liquid sodium and internal structures of the reactor). A short parametric study shows the influence of the porosity and the structure shape on the pressure wave impacting the shock tube bottom. These results were used to simulate numerically the HCDA mechanical effects in a small scale reactor mock-up. (author)

  19. Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors. Publishable Final Activity Report

    Energy Technology Data Exchange (ETDEWEB)

    Kuijper, J.C., E-mail: kuijper@nrg.eu [Nuclear Research and Consultancy Group (NRG), Petten (Netherlands); Somers, J.; Van Den Durpel, L.; Chauvet, V.; Cerullo, N.; Cetnar, J.; Abram, T.; Bakker, K.; Bomboni, E.; Bernnat, W.; Domanska, J.G.; Girardi, E.; De Haas, J.B.M.; Hesketh, K.; Hiernaut, J.P.; Hossain, K.; Jonnet, J.; Kim, Y.; Kloosterman, J.L.; Kopec, M.; Murgatroyd, J.; Millington, D.; Lecarpentier, D.; Lomonaco, G.; McEachern, D.; Meier, A.; Mignanelli, M.; Nabielek, H.; Oppe, J.; Petrov, B.Y.; Pohl, C.; Ruetten, H.J.; Schihab, S.; Toury, G.; Trakas, C.; Venneri, F.; Verfondern, K.; Werner, H.; Wiss, T.; Zakova, J.

    2010-11-15

    The PUMA project -the acronym stands for 'Plutonium and Minor Actinide Management in Thermal High-Temperature Gas-Cooled Reactors'- was a Specific Targeted Research Project (STREP) within the EURATOM 6th Framework Program (EU FP6). The PUMA project ran from September 1, 2006, until August 31, 2009, and was executed by a consortium of 14 European partner organisations and one from the USA. This report serves 2 purposes. It is both the 'Publishable Final Activity Report' and the 'Final (Summary) Report', describing, per Work Package, the specific objectives, research activities, main conclusions, recommendations and supporting documents. PUMA's main objective was to investigate the possibilities for the utilisation and transmutation of plutonium and especially minor actinides in contemporary and future (high temperature) gas-cooled reactor designs, which are promising tools for improving the sustainability of the nuclear fuel cycle. This contributes to the reduction of Pu and MA stockpiles, and also to the development of safe and sustainable reactors for CO{sub 2}-free energy generation. The PUMA project has assessed the impact of the introduction of Pu/MA-burning HTRs at three levels: fuel and fuel performance (modelling), reactor (transmutation performance and safety) and reactor/fuel cycle facility park. Earlier projects already indicated favourable characteristics of HTRs with respect to Pu burning. So, core physics of Pu/MA fuel cycles for HTRs has been investigated to study the CP fuel and reactor characteristics and to assure nuclear stability of a Pu/MA HTR core, under both normal and abnormal operating conditions. The starting point of this investigation comprised the two main contemporary HTR designs, viz. the pebble-bed type HTR, represented by the South-African PBMR, and hexagonal block type HTR, represented by the GT-MHR. The results (once again) demonstrate the flexibility of the contemporary (and near future) HTR

  20. On the Safety and Performance Demonstration Tests of Prototype Gen-IV Sodium-Cooled Fast Reactor and Validation and Verification of Computational Codes

    Directory of Open Access Journals (Sweden)

    Jong-Bum Kim

    2016-10-01

    Full Text Available The design of Prototype Gen-IV Sodium-Cooled Fast Reactor (PGSFR has been developed and the validation and verification (V&V activities to demonstrate the system performance and safety are in progress. In this paper, the current status of test activities is described briefly and significant results are discussed. The large-scale sodium thermal-hydraulic test program, Sodium Test Loop for Safety Simulation and Assessment-1 (STELLA-1, produced satisfactory results, which were used for the computer codes V&V, and the performance test results of the model pump in sodium showed good agreement with those in water. The second phase of the STELLA program with the integral effect tests facility, STELLA-2, is in the detailed design stage of the design process. The sodium thermal-hydraulic experiment loop for finned-tube sodium-to-air heat exchanger performance test, the intermediate heat exchanger test facility, and the test facility for the reactor flow distribution are underway. Flow characteristics test in subchannels of a wire-wrapped rod bundle has been carried out for safety analysis in the core and the dynamic characteristic test of upper internal structure has been performed for the seismic analysis model for the PGSFR. The performance tests for control rod assemblies (CRAs have been conducted for control rod drive mechanism driving parts and drop tests of the CRA under scram condition were performed. Finally, three types of inspection sensors under development for the safe operation of the PGSFR were explained with significant results.

  1. Advanced Gas Cooled Nuclear Reactor Materials Evaluation and Development Program. Progress report, April 1, 1980-June 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-11-14

    Objectives of this program are to evaluate candidate alloys for Very High Temperature Reactor (VHTR) Nuclear Process Heat (NPH) and Direct Cycle Helium Turbine (DCHT) applications, in terms of the effect of simulated reactor primary coolant (helium containing small amounts of various other gases), high temperatures, and long time exposures, on the mechanical properties and structural and surface stability of selected candidate alloys. A second objective is to select and recommend materials for future test facilities and more extensive qualification programs. Work covered in this report includes the activities associated with the status of the simulated reactor helium supply system, testing equipment and gas chemistry analysis instrumentation and equipment. The progress in the screening test program is described; this includes: screening creep results and metallographic analysis for materials thermally exposed or tested at 750, 850 and 950/sup 0/C. The initiation of air creep-rupture testing in the intensive screening test program is discussed. In addition, the status of the data management system is described.

  2. Reliability Assessment of 2400 MWth Gas-Cooled Fast Reactor Natural Circulation Decay Heat Removal in Pressurized Situations

    Directory of Open Access Journals (Sweden)

    C. Bassi

    2008-01-01

    Full Text Available As the 2400 MWth gas-cooled fast reactor concept makes use of passive safety features in combination with active safety systems, the question of natural circulation decay heat removal (NCDHR reliability and performance assessment into the ongoing probabilistic safety assessment in support to the reactor design, named “probabilistic engineering assessment” (PEA, constitutes a challenge. Within the 5th Framework Program for Research and Development (FPRD of the European Community, a methodology has been developed to evaluate the reliability of passive systems characterized by a moving fluid and whose operation is based on physical principles, such as the natural circulation. This reliability method for passive systems (RMPSs is based on uncertainties propagation into thermal-hydraulic (T-H calculations. The aim of this exercise is finally to determine the performance reliability of the DHR system operating in a “passive” mode, taking into account the uncertainties of parameters retained for thermal-hydraulical calculations performed with the CATHARE 2 code. According to the PEA preliminary results, exhibiting the weight of pressurized scenarios (i.e., with intact primary circuit boundary for the core damage frequency (CDF, the RMPS exercise is first focusing on the NCDHR performance at these T-H conditions.

  3. Methodology for sodium fire vulnerability assessment of sodium cooled fast reactor based on the Monte-Carlo principle

    Energy Technology Data Exchange (ETDEWEB)

    Song, Wei [Nuclear and Radiation Safety Center, P. O. Box 8088, Beijing (China); Wu, Yuanyu [ITER Organization, Route de Vinon-sur-Verdon, 13115 Saint-Paul-lès-Durance (France); Hu, Wenjun [China Institute of Atomic Energy, P. O. Box 275(34), Beijing (China); Zuo, Jiaxu, E-mail: zuojiaxu@chinansc.cn [Nuclear and Radiation Safety Center, P. O. Box 8088, Beijing (China)

    2015-11-15

    Highlights: • Monte-Carlo principle coupling with fire dynamic code is adopted to perform sodium fire vulnerability assessment. • The method can be used to calculate the failure probability of sodium fire scenarios. • A calculation example and results are given to illustrate the feasibility of the methodology. • Some critical parameters and experience are shared. - Abstract: Sodium fire is a typical and distinctive hazard in sodium cooled fast reactors, which is significant for nuclear safety. In this paper, a method of sodium fire vulnerability assessment based on the Monte-Carlo principle was introduced, which could be used to calculate the probabilities of every failure mode in sodium fire scenarios. After that, the sodium fire scenario vulnerability assessment of primary cold trap room of China Experimental Fast Reactor was performed to illustrate the feasibility of the methodology. The calculation result of the example shows that the conditional failure probability of key cable is 23.6% in the sodium fire scenario which is caused by continuous sodium leakage because of the isolation device failure, but the wall temperature, the room pressure and the aerosol discharge mass are all lower than the safety limits.

  4. On0Line Fuel Failure Monitor for Fuel Testing and Monitoring of Gas Cooled Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ayman I. Hawari; Mohamed A. Bourham

    2010-04-22

    IVery High Temperature Reactors (VHTR) utilize the TRISO microsphere as the fundamental fuel unit in the core. The TRISO microsphere (~ 1- mm diameter) is composed of a UO2 kernel surrounded by a porous pyrolytic graphite buffer, an inner pyrolytic graphite layer, a silicon carbide (SiC) coating, and an outer pyrolytic graphite layer. The U-235 enrichment of the fuel is expected to range from 4% – 10% (higher enrichments are also being considered). The layer/coating system that surrounds the UO2 kernel acts as the containment and main barrier against the environmental release of radioactivity. To understand better the behavior of this fuel under in-core conditions (e.g., high temperature, intense fast neutron flux, etc.), the US Department of Energy (DOE) is launching a fuel testing program that will take place at the Advanced Test Reactor (ATR) located at Idaho National Laboratory (INL). During this project North Carolina State University (NCSU) researchers will collaborate with INL staff for establishing an optimized system for fuel monitoring for the ATR tests. In addition, it is expected that the developed system and methods will be of general use for fuel failure monitoring in gas cooled VHTRs.

  5. An Artificial Neural Network Compensated Output Feedback Power-Level Control for Modular High Temperature Gas-Cooled Reactors

    Directory of Open Access Journals (Sweden)

    Zhe Dong

    2014-02-01

    Full Text Available Small modular reactors (SMRs could be beneficial in providing electricity power safely and also be viable for applications such as seawater desalination and heat production. Due to its inherent safety features, the modular high temperature gas-cooled reactor (MHTGR has been seen as one of the best candidates for building SMR-based nuclear power plants. Since the MHTGR dynamics display high nonlinearity and parameter uncertainty, it is necessary to develop a nonlinear adaptive power-level control law which is not only beneficial to the safe, stable, efficient and autonomous operation of the MHTGR, but also easy to implement practically. In this paper, based on the concept of shifted-ectropy and the physically-based control design approach, it is proved theoretically that the simple proportional-differential (PD output-feedback power-level control can provide asymptotic closed-loop stability. Then, based on the strong approximation capability of the multi-layer perceptron (MLP artificial neural network (ANN, a compensator is established to suppress the negative influence caused by system parameter uncertainty. It is also proved that the MLP-compensated PD power-level control law constituted by an experientially-tuned PD regulator and this MLP-based compensator can guarantee bounded closed-loop stability. Numerical simulation results not only verify the theoretical results, but also illustrate the high performance of this MLP-compensated PD power-level controller in suppressing the oscillation of process variables caused by system parameter uncertainty.

  6. Evaluation of radiation heat transfer in porous medial: Application for a pebble bed modular reactor cooled by CO2 gas

    Directory of Open Access Journals (Sweden)

    Sidi-Ali Kamel

    2013-01-01

    Full Text Available This work analyses the contribution of radiation heat transfer in the cooling of a pebble bed modular reactor. The mathematical model, developed for a porous medium, is based on a set of equations applied to an annular geometry. Previous major works dealing with the subject have considered the forced convection mode and often did not take into account the radiation heat transfer. In this work, only free convection and radiation heat transfer are considered. This can occur during the removal of residual heat after shutdown or during an emergency situation. In order to derive the governing equations of radiation heat transfer, a steady-state in an isotropic and emissive porous medium (CO2 is considered. The obtained system of equations is written in a dimensionless form and then solved. In order to evaluate the effect of radiation heat transfer on the total heat removed, an analytical method for solving the system of equations is used. The results allow quantifying both radiation and free convection heat transfer. For the studied situation, they show that, in a pebble bed modular reactor, more than 70% of heat is removed by radiation heat transfer when CO2 is used as the coolant gas.

  7. Development of Advanced 9Cr Ferritic-Martensitic Steels and Austenitic Stainless Steels for Sodium-Cooled Fast Reactor