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Sample records for air cooled reactors

  1. Temperature distribution in graphite during annealing in air cooled reactors

    International Nuclear Information System (INIS)

    A model for the evaluation temperature distributions in graphite during annealing operation in graphite. Moderated an-cooled reactors, is presented. One single channel and one dimension for air and graphite were considered. A numerical method based on finite control volumes was used for partioning the mathematical equations. The problem solution involves the use of unsteady equations of mass, momentum and energy conservation for air, and energy conservation for graphite. The source term was considered as stored energy release during annealing for describing energy conservation in the graphite. The coupling of energy conservation equations in air and graphite is performed by the heat transfer term betwen air and graphite. The results agree with experimental data. A sensitivity analysis shown that the termal conductivity of graphite and the maximum inlet channel temperature have great effect on the maximum temperature reached in graphite during the annealing. (author)

  2. 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.

  3. Preliminary Design of KAIST Micro Modular Reactor with Dry Air Cooling

    Energy Technology Data Exchange (ETDEWEB)

    Baik, Seung Joon; Bae, Seong Jun; Kim, Seong Gu; Lee, Jeong Ik [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    KAIST research team recently proposed a Micro Modular Reactor (MMR) concept which integrates power conversion unit (PCU) with the reactor core in a single module. Using supercritical CO{sub 2} as a working fluid of cycle can achieve physically compact size due to small turbomachinery and heat exchangers. The objective of this project is to develop a concept that can operate at isolated area. The design focuses especially on the operation in the inland area where cooling water is insufficient. Thus, in this paper the potential for dry air cooling of the proposed reactor will be examined by sizing the cooling system with preliminary approach. The KAIST MMR is a recently proposed concept of futuristic SMR. The MMR size is being determined to be transportable with land transportation. Special attention is given to the MMR design on the dry cooling, which the cooling system does not depend on water. With appropriately designed air cooling heat exchanger, the MMR can operate autonomously. Two types of air cooling methods are suggested. One is using fan and the other is utilizing cooling tower for the air flow. With fan type air cooling method it consumes about 0.6% of generated electricity from the nuclear reactor. Cooling tower occupies an area of 227 m{sup 2} and 59.6 m in height. This design is just a preliminary estimation of the dry cooling method, and therefore more detailed and optimal design will be followed in the next phase.

  4. 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.

  5. STUDY ON AIR INGRESS MITIGATION METHODS IN THE VERY HIGH TEMPERATURE GAS COOLED REACTOR (VHTR)

    Energy Technology Data Exchange (ETDEWEB)

    Chang H. Oh

    2011-03-01

    An air-ingress accident followed by a pipe break is considered as a critical event for a very high temperature gas-cooled reactor (VHTR). Following helium depressurization, it is anticipated that unless countermeasures are taken, air will enter the core through the break leading to oxidation of the in-core graphite structure. Thus, without mitigation features, this accident might lead to severe exothermic chemical reactions of graphite and oxygen. Under extreme circumstances, a loss of core structural integrity may occur along with excessive release of radiological inventory. Idaho National Laboratory under the auspices of the U.S. Department of Energy is performing research and development (R&D) that focuses on key phenomena important during challenging scenarios that may occur in the VHTR. Phenomena Identification and Ranking Table (PIRT) studies to date have identified the air ingress event, following on the heels of a VHTR depressurization, as very important (Oh et al. 2006, Schultz et al. 2006). Consequently, the development of advanced air ingress-related models and verification and validation (V&V) requirements are part of the experimental validation plan. This paper discusses about various air-ingress mitigation concepts applicable for the VHTRs. The study begins with identifying important factors (or phenomena) associated with the air-ingress accident by using a root-cause analysis. By preventing main causes of the important events identified in the root-cause diagram, the basic air-ingress mitigation ideas can be conceptually derived. The main concepts include (1) preventing structural degradation of graphite supporters; (2) preventing local stress concentration in the supporter; (3) preventing graphite oxidation; (4) preventing air ingress; (5) preventing density gradient driven flow; (4) preventing fluid density gradient; (5) preventing fluid temperature gradient; (6) preventing high temperature. Based on the basic concepts listed above, various air

  6. Performance Estimation of Supercritical Co2 Micro Modular Reactor (MMR) for Varying Cooling Air Temperature

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Yoonhan; Kim, Seong Gu; Cho, Seong Kuk; Lee, Jeong Ik [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2015-05-15

    A Small Modular Reactor (SMR) receives interests for the various application such as electricity co-generation, small-scale power generation, seawater desalination, district heating and propulsion. As a part of SMR development, supercritical CO2 Micro Modular Reactor (MMR) of 36.2MWth in power is under development by the KAIST research team. To enhance the mobility, the entire system including the power conversion system is designed for the full modularization. Based on the preliminary design, the thermal efficiency is 31.5% when CO2 is sufficiently cooled to the design temperature. A supercritical CO2 MMR is designed to supply electricity to the remote regions. The ambient temperature of the area can influence the compressor inlet temperature as the reactor is cooled with the atmospheric air. To estimate the S-CO2 cycle performance for various environmental conditions, A quasi-static analysis code is developed. For the off design performance of S-CO2 turbomachineries, the experimental result of Sandia National Lab (SNL) is utilized.

  7. 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.

  8. Cooling system for reactor container

    International Nuclear Information System (INIS)

    Purpose: To effectively cool a reactor container upon reactor shutdown with no intrusion of metal corrosion products in coolants into the main steam pipe in a BWR type reactor. Constitution: A clean up system comprising a pipeway, a recycling pump, a non-regenerative heat exchanger and a primary coolant purifier and a regenerative heat exchanger is provided branched from a residual heat removing system and the clean up system is connected by way of a valve to a feedwater pipeway, as well as connected by way of the pipeway to the main steam pipeway at the midway of two main steam separation valves outside of the reactor container. This enables to prevent metal corrosion products floating on the surface of reactor water from introducing into the main steam pipe when the pressure vessel is filled with water. Then, since the pressure vessel is filled with primary coolants, the pressure vessel can be cooled uniformly in a short time. (Ikeda, J.)

  9. Pressure drop and heat transfer in the sodium to air heat exchanger tube banks on advanced sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    A numerical study was performed to investigate the thermal and hydraulic characteristics and build up design model of the AHX (sodium-to-air heat exchanger) unit of a sodium-cooled fast reactor. Helical-coiled tube banks in the AHX were modeled as porous media and simulated heat and momentum transfer. Two-dimensional flow characteristic appeared at the most region of AHX annulus. Pressure drop and heat transfer coefficient for rectangular, parallelogram and staggered tube banks as the main components of the AHX were evaluated and compared with Zhukauskas empirical correlations. (author)

  10. Dynamic simulation of the air-cooled decay heat removal system of the German KNK-II experimental breeder reactor

    International Nuclear Information System (INIS)

    A Dump Heat Exchanger and associated feedback control system models for decay heat removal in the German KNK-II experimental fast breeder reactor are presented. The purpose of the controller is to minimize temperature variations in the circuits and, hence, to prevent thermal shocks in the structures. The basic models for the DHX include the sodium-air thermodynamics and hydraulics, as well as a control system. Valve control models for the primary and intermediate sodium flow regulation during post shutdown conditions are also presented. These models have been interfaced with the SSC-L code. Typical results of sample transients are discussed

  11. Gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    The gas temperature of a hot gas loop in gas-cooled nuclear reactor plants shall be able to be modified without influencing the gas temperature of the other loops. If necessary, it should be possible to stop the loop. This is possible by means of a mixer which is places below the heat absorbing component in the hot channel and which is connected to a cold gas line. (orig.)

  12. Gas-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Almost all the R D works of gas-cooled fast breeder reactor in the world were terminated at the end of the year 1980. In order to show that the R D termination was not due to technical difficulties of the reactor itself, the present paper describes the reactor plant concept, reactor performances, safety, economics and fuel cycle characteristics of the reactor, and also describes the reactor technologies developed so far, technological problems remained to be solved and planned development schedules of the reactor. (author)

  13. Natural Flow Air Cooled Photovoltaics

    Science.gov (United States)

    Tanagnostopoulos, Y.; Themelis, P.

    2010-01-01

    Our experimental study aims to investigate the improvement in the electrical performance of a photovoltaic installation on buildings through cooling of the photovoltaic panels with natural air flow. Our experimental study aims to investigate the improvement in the electrical performance of a photovoltaic installation on buildings through cooling of the photovoltaic panels with natural air flow. We performed experiments using a prototype based on three silicon photovoltaic modules placed in series to simulate a typical sloping building roof with photovoltaic installation. In this system the air flows through a channel on the rear side of PV panels. The potential for increasing the heat exchange from the photovoltaic panel to the circulating air by the addition of a thin metal sheet (TMS) in the middle of air channel or metal fins (FIN) along the air duct was examined. The operation of the device was studied with the air duct closed tightly to avoid air circulation (CLOSED) and the air duct open (REF), with the thin metal sheet (TMS) and with metal fins (FIN). In each case the experiments were performed under sunlight and the operating parameters of the experimental device determining the electrical and thermal performance of the system were observed and recorded during a whole day and for several days. We collected the data and form PV panels from the comparative diagrams of the experimental results regarding the temperature of solar cells, the electrical efficiency of the installation, the temperature of the back wall of the air duct and the temperature difference in the entrance and exit of the air duct. The comparative results from the measurements determine the improvement in electrical performance of the photovoltaic cells because of the reduction of their temperature, which is achieved by the naturally circulating air.

  14. Ultrasonic methodology measurement of two-phase (air-water) flows in cooling systems of nuclear reactor

    International Nuclear Information System (INIS)

    An ultrasonic methodology is proposed for the measurement for two phase (air-water) flow parameters. Ultrasonic backscattered signals were used to analyze the following parameters: average number of bubbles, interfacial area and void fraction. The results show a strong correlation between the parameters and the ultrasonic power signal obtained. (author)

  15. Gas-cooled reactors and their applications

    International Nuclear Information System (INIS)

    The purpose of the meeting was to review and discuss the current status and recent progress made in the technology and design of gas-cooled reactors and their application for electricity generation, process steam and process heat production. The meeting was attended by more than 200 participants from 25 countries and International Organizations presenting 34 papers. The technical part of the meeting was subdivided into 7 sessions: A. Overview of the Status of Gas-Cooled Reactors and Their Prospects (2 papers); B. Experience with Gas-Cooled Reactors (5 papers); C. Description of Current GCR Plant Designs (10 papers); D. Safety Aspects (4 papers); E. Gas-Cooled Reactor Applications (3 papers); F. Gas-Cooled Reactor Technology (6 papers); G. User's Perspectives on Gas-Cooled Reactors (4 papers). At the end of the meeting a round table discussion was organized in order to summarize the meeting and to make recommendations for future activities. A separate abstract was prepared for each of the 34 presentations of this meeting. Refs, figs and tabs

  16. French activities on gas cooled reactors

    International Nuclear Information System (INIS)

    The gas cooled reactor programme in France originally consisted of eight Natural Uranium Graphite Gas Cooled Reactors (UNGG). These eight units, which are now permanently shutdown, represented a combined net electrical power of 2,375 MW and a total operational history of 163 years. Studies related to these reactors concern monitoring and dismantling of decommissioned facilities, including the development of methods for dismantling. France has been monitoring the development of HTRs throughout the world since 1979, when it halted its own HTR R and D programme. France actively participates in three CRPs set up by the IAEA. (author). 1 tab

  17. 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

  18. Materials for advanced water cooled reactors

    International Nuclear Information System (INIS)

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

  19. Liquid-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Hydrogen can be added to nuclear reactors with a liquid hydrogen-containing coolant on the suction side of a high pressure pump in the purification system. According to the invention this is performed by means of a liquid jet condenser which uses the coolant as liquid and which is preferably charged from the pressure side of the high pressure pump and conveys the liquid to a mixer connected in series with the high pressure pump. The invention is to be used especially in pressurized water reactors. (orig.)

  20. 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.

  1. Thermohydraulic relationships for advanced water cooled reactors

    International Nuclear Information System (INIS)

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

  2. Passive cooling in modern nuclear reactors

    International Nuclear Information System (INIS)

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

  3. Neutronic of heterogenous gas cooled reactors

    International Nuclear Information System (INIS)

    At present, one of the main technical features of the advanced gas cooled reactor under development is its fuel element concept, which implies a neutronic homogeneous design, thus requiring higher enrichment compared with present commercial nuclear power plants.In this work a neutronic heterogeneous gas cooled reactor design is analyzed by studying the neutronic design of the Advanced Gas cooled Reactor (AGR), a low enrichment, gas cooled and graphite moderated nuclear power plant.A search of merit figures (some neutronic parameter, characteristic dimension, or a mixture of both) which are important and have been optimized during the reactor design stage is been done, to aim to comprise how a gas heterogeneous reactor is been design, given that semi-infinity arrangement criteria of rods in LWRs and clusters in HWRs can t be applied for a solid moderator and a gas refrigerator.The WIMS code for neutronic cell calculations is been utilized to model the AGR fuel cell and to calculate neutronic parameters such as the multiplication factor and the pick factor, as function of the fuel burnup.Also calculation is been done for various nucleus characteristic dimensions values (fuel pin radius, fuel channel pitch) and neutronic parameters (such as fuel enrichment), around the design established parameters values.A fuel cycle cost analysis is carried out according to the reactor in study, and the enrichment effect over it is been studied.Finally, a thermal stability analysis is been done, in subcritical condition and at power level, to study this reactor characteristic reactivity coefficients.Present results shows (considering the approximation used) a first set of neutronic design figures of merit consistent with the AGR design.

  4. Mixed oxide fuel for water cooled reactors

    International Nuclear Information System (INIS)

    The problems connected with introduction of plutonium extracted from spent fuels of operating NPPs into water cooled reactor fuel cycle are considered. The trends in formation of the World market of mixed fuel are illustrated taking as examples Great Britain and Japan

  5. Coolant technology of water cooled reactors. V. 1: Chemistry of primary coolant in water cooled reactors

    International Nuclear Information System (INIS)

    This report is a summary of the work performed within the framework of the Coordinated Research Programme on Investigations on Water Chemistry Control and Coolant Interaction with Fuel and Primary Circuit Materials in Water Cooled Power Reactors organized by the IAEA and carried out from 1987 to 1991. It is the continuation of a programme entitled Reactor Water Chemistry Relevant to Coolant-Cladding Interaction (IAEA-TECDOC-429), which ran from 1981 to 1986. Subsequent meetings resulted in the title of the programme being changed to Coolant Technology of Water Cooled Reactors. The results of this Coordinated Research Programme are published in four volumes with an overview in the Technical Reports Series. The titles of the volumes are: Volume 1: Chemistry of Primary Coolant in Water Cooled Reactors; Volume 2: Corrosion in the Primary Coolant Systems of Water Cooled Reactors; Volume 3: Activity Transport Mechanisms in Water Cooled Reactors; Volume 4: Decontamination of Water Cooled Reactors. These publications should be of interest to experts in water chemistry at nuclear power plants, experts in engineering, fuel designers, research and development institutes active in the field and to consultants to these organizations. Refs, figs and tabs

  6. Startup of the FFTF sodium cooled reactor

    International Nuclear Information System (INIS)

    The Fast Flux Test Facility (FFTF), located on the Department of Energy (DOE) Hanford Reservation near Richland, Washington, is a 3 Loop 400 MW(t) sodium cooled fast reactor with a primary mission to test fuels and materials for development of the Liquid Metal Fast Breeder Reactor (LMFBR). Bringing FFTF to a condition to accomplish this mission is the goal of the Acceptance Test Program (ATP). This program was the mechanism for achieving startup of the FFTF. Highlights of the ATP involving the system inerting, liquid metal and inerted cell testing and initial ascent to full power are discussed

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  8. Analysis of two phase natural circulation flow in the reactor cavity under external reactor vessel cooling

    International Nuclear Information System (INIS)

    As part of a study on a two-phase natural circulation flow between the outer reactor vessel and the insulation material in the reactor cavity under an external reactor vessel cooling of APR (Advanced Power Reactor) 1400, a K-HERMES-HALF (Hydraulic Evaluation of Reactor cooling Mechanism by External Self-induced flow-HALF scale) experiment was performed at KAERI (Korea Atomic Energy Research Institute) using an air injection method. This experiment was analyzed to verify and evaluate the experimental results using the RELAP5/MOD3 computer code. In addition, the geometry scaling on full height & full sector, and a material scaling between air-water and steam-water two phase natural circulation flow, have been performed for an application of the experimental results to an actual APR1400. The RELAP5/MOD3 results on the water circulation mass flow rate are very similar to the experimental results, in general. The water circulation mass flow rate of the full height & full sector case is approximately 7.6-times higher than that of the K-HERMEL-HALF case. The water circulation mass flow rate of the air injection case is 20-50 % higher than that of the steam injection case at 20 % of the injection rate. (author)

  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. Bacterial pathogens in a reactor cooling reservoir

    International Nuclear Information System (INIS)

    The results of the sampling in both Par Pond and Clark Hill Reservoir are given. The frequency of isolation is a qualitative parameter which indicates how often the specified bacterium was isolated from each habitat. Initial scoping experiments demonstrated that a wider variety of pathogenic bacteria occur in Par Pond than in Clark Hill Reservoir. Such findings are interesting because Par Pond does not receive any human wastes directly, yet bacteria generally associated with human wastes are more frequently isolated from Par Pond. Previous studies have demonstrated that certain non-spore-forming enteric bacteria do not survive the intense heat associated with the cooling water when the reactor is operating. However, even when the reactor is not operating, cooling water, consisting of 10% makeup water from Savannah River, continues to flow into Par Pond. This flow provides a source of bacteria which inoculate Par Pond. Once the reactor is again operating, these same bacteria appear to be able to survive and grow within the Par Pond system. Thus, Par Pond and the associated lakes and canals of the Par Pond system provide a pool of pathogens that normally would not survive in natural waters

  11. Design codes for gas cooled reactor components

    International Nuclear Information System (INIS)

    High-temperature gas-cooled reactor (HTGR) plants have been under development for about 30 years and experimental and prototype plants have been operated. The main line of development has been electricity generation based on the steam cycle. In addition the potential for high primary coolant temperature has resulted in research and development programmes for advanced applications including the direct cycle gas turbine and process heat applications. In order to compare results of the design techniques of various countries for high temperature reactor components, the IAEA established a Co-ordinated Research Programme (CRP) on Design Codes for Gas-Cooled Reactor Components. The Federal Republic of Germany, Japan, Switzerland and the USSR participated in this Co-ordinated Research Programme. Within the frame of this CRP a benchmark problem was established for the design of the hot steam header of the steam generator of an HTGR for electricity generation. This report presents the results of that effort. The publication also contains 5 reports presented by the participants. A separate abstract was prepared for each of these reports. Refs, figs and tabs

  12. Lessons Learned From Gen I Carbon Dioxide Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    David E. Shropshire

    2004-04-01

    This paper provides a review of early gas cooled reactors including the Magnox reactors originating in the United Kingdom and the subsequent development of the Advanced Gas-cooled Reactors (AGR). These early gas cooled reactors shared a common coolant medium, namely carbon dioxide (CO2). A framework of information is provided about these early reactors and identifies unique problems/opportunities associated with use of CO2 as a coolant. Reactor designers successfully rose to these challenges. After years of successful use of the CO2 gas cooled reactors in Europe, the succeeding generation of reactors, called the High Temperature Gas Reactors (HTGR), were designed with Helium gas as the coolant. Again, in the 21st century, with the latest reactor designs under investigation in Generation IV, there is a revived interest in developing Gas Cooled Fast Reactors that use CO2 as the reactor coolant. This paper provides a historical perspective on the 52 CO2 reactors and the reactor programs that developed them. The Magnox and AGR design features and safety characteristics were reviewed, as well as the technologies associated with fuel storage, reprocessing, and disposal. Lessons-learned from these programs are noted to benefit the designs of future generations of gas cooled nuclear reactors.

  13. 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

  14. LFR "Lead-Cooled Fast Reactor"

    Energy Technology Data Exchange (ETDEWEB)

    Cinotti, L; Fazio, C; Knebel, J; Monti, S; Abderrahim, H A; Smith, C; Suh, K

    2006-05-11

    The main purpose of this paper is to present the current status of development of the Lead-cooled Fast Reactor (LFR) in Generation IV (GEN IV), including the European contribution, to identify needed R&D and to present the corresponding GEN IV International Forum (GIF) R&D plan [1] to support the future development and deployment of lead-cooled fast reactors. The approach of the GIF plan is to consider the research priorities of each member country in proposing an integrated, coordinated R&D program to achieve common objectives, while avoiding duplication of effort. The integrated plan recognizes two principal technology tracks: (1) a small, transportable system of 10-100 MWe size that features a very long refuelling interval, and (2) a larger-sized system rated at about 600 MWe, intended for central station power generation. This paper provides some details of the important European contributions to the development of the LFR. Sixteen European organizations have, in fact, taken the initiative to present to the European Commission the proposal for a Specific Targeted Research and Training Project (STREP) devoted to the development of a European Lead-cooled System, known as the ELSY project; two additional organizations from the US and Korea have joined the project. Consequently, ELSY will constitute the reference system for the large lead-cooled reactor of GEN IV. The ELSY project aims to demonstrate the feasibility of designing a competitive and safe fast power reactor based on simple technical engineered features that achieves all of the GEN IV goals and gives assurance of investment protection. As far as new technology development is concerned, only a limited amount of R&D will be conducted in the initial phase of the ELSY project since the first priority is to define the design guidelines before launching a larger and expensive specific R&D program. In addition, the ELSY project is expected to benefit greatly from ongoing lead and lead-alloy technology

  15. Electrochemistry of Water-Cooled Nuclear Reactors

    International Nuclear Information System (INIS)

    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

  16. 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.

  17. Status of and prospects for gas-cooled reactors

    International Nuclear Information System (INIS)

    The IAEA International Working Group on Gas-Cooled Reactors (IWGGCR) (see Annex I), which was established in 1978, recommended to the Agency that a report be prepared in order to provide an up-to-date summary of gas-cooled reactor technology. The present Technical Report is based mainly on submissions of Member Countries of the IWGGCR and consists of four main sections. Beside some general information about the gas-cooled reactor line, section 1 contains a description of the incentives for the development and deployment of gas-cooled reactors in various Agency Member States. These include both electricity generation and process steam and process heat production for various branches of industry. The historical development of gas-cooled reactors is reviewed in section 2. In this section information is provided on how, when and why gas-cooled reactors have been developed in various Agency Member States and, in addition, a detailed description of the different gas-cooled reactor lines is presented. Section 3 contains information about the technical status of gas-cooled reactors and their applications. Gas-cooled reactors that are under design or construction or in operation are listed and shortly described, together with an outlook for future reactor designs. In this section the various applications for gas-cooled reactors are described in detail. These include both electricity generation and process steam and process heat production. The last section (section 4) is entitled ''Special features of gas-cooled reactors'' and contains information about the technical performance, fuel utilization, safety characteristics and environmental impact, such as radiation exposure and heat rejection

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

    International Nuclear Information System (INIS)

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

  19. Status of national gas cooled reactor programmes

    International Nuclear Information System (INIS)

    This report has been compiled as a central source of summary-level information on the present status of High Temperature Gas-Cooled Reactor (HTGR) programmes in the world and on future plans for the continued development and deployment of HTGRs. Most of the information concerns the programmes in the United States, Germany, Japan and the Soviet Union, countries that have had large programmes related to HTGR technology for several years. Summary-level information is also provided in the report on HTGR-related activities in several other countries who either have an increasing interest in the technology and/or who are performing some development efforts related to HTGR technology. The report contains a summary-level update on the MAGNOX and AGR programmes. This is the twelfth issue of the document, the first of which was issued in March, 1979. The report has been prepared in the IAEA Nuclear Power Technology Development Section. Figs and tabs

  20. Seismic behaviour of gas cooled reactor components

    International Nuclear Information System (INIS)

    On invitation of the French Government the Specialists' Meeting on the Seismic Behaviour of Gas-Cooled Reactor Components was held at Gif-sur-Yvette, 14-16 November 1989. This was the second Specialists' Meeting on the general subject of gas-cooled reactor seismic design. There were 27 participants from France, the Federal Republic of Germany, Israel, Japan, Spain, Switzerland, the United Kingdom, the Soviet Union, the United States, the CEC and IAEA took the opportunity to present and discuss a total of 16 papers reflecting the state of the art of gained experiences in the field of their seismic qualification approach, seismic analysis methods and of the capabilities of various facilities used to qualify components and verify analytical methods. Since the first meeting, the sophistication and expanded capabilities of both the seismic analytical methods and the test facilities are apparent. The two main methods for seismic analysis, the impedance method and the finite element method, have been computer-programmed in several countries with the capability of each of the codes dependent on the computer capability. The correlations between calculation and tests are dependent on input assumptions such as boundary conditions, soil parameters and various interactions between the soil, the buildings and the contained equipment. The ability to adjust these parameters and match experimental results with calculations was displayed in several of the papers. The expanded capability of some of the new test facilities was graphically displayed by the description of the SAMSON vibration test facility at Juelich, FRG, capable of dynamically testing specimens weighing up to 25 tonnes, and the TAMARIS facility at the CEA laboratories in Gif-sur-Yvette where the largest table is capable of testing specimens weighing up to 100 tonnes. The proceedings of this meeting contain all 16 presented papers. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  1. 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...

  2. A Blast of Cool Air

    Science.gov (United States)

    2000-01-01

    Unable to solve their engineering problem with a rotor in their Orbital Vane product, DynEco Corporation turned to Kennedy Space Center for help. KSC engineers determined that the compressor rotor was causing a large concentration of stress, which led to cracking and instant rotor failure. NASA redesigned the lubrication system, which allowed the company to move forward with its compressor that has no rubbing parts. The Orbital Vane is a refrigerant compressor suitable for mobile air conditioning and refrigeration.

  3. Pre-Analysis of Triga Mark II Reactor Cooling System

    OpenAIRE

    AKAY, Orhan Erdal

    2012-01-01

    In this study, work of the reactor cooling system is divided into two time zone. The second cooling circuit has been that the conditions required operating. Cooling system which is the center of the heat exchanger total heat transfer coefficient correlations were calculated using the theoretical. The design values were compared with results obtained by calculation.

  4. Development of studies on helium cooled fast reactors

    International Nuclear Information System (INIS)

    A necessity is shown of developing breeders with high reproductive properties. Helium cooled fast reactor is considered. The reactor performances, heating circuit with the use of a steam turbine unit in the secondary circuit is outlined. The reactor design and fuel assemblies are described

  5. 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.

  6. 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.

  7. 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...

  8. 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...

  9. for an Internal Cooling Air System

    Directory of Open Access Journals (Sweden)

    Tadaharu Kishibe

    2000-01-01

    Full Text Available The swirling flow field in an internal cooling air system in which the fluid passes through an inducer, a hollow turbine shaft, and a cavity between two disks (referred to as a wheel space is solved using computational fluid dynamics and the pressure fluctuations on the hollow shaft wall surface are measured.

  10. 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.

  11. Design Considerations for Economically Competitive Sodium Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hongbin Zhang; Haihua Zhao

    2009-05-01

    The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phénix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design.

  12. 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. .

  13. Seismic stability of VGM type high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    The main principles of the design provision of high temperature gas cooled VGM reactors seismic stability and the results of calculations, performed by linear-spectral method are presented. (author). 1 ref., 10 figs

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

    OpenAIRE

    V. V. Birangane; A.M.Patil

    2014-01-01

    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 ...

  15. 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

    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...

  16. 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.

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

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1990-01-01

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

  18. Passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1989-01-01

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

  19. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

    The research and development work in the field of core thermal-hydraulics, experimental and analytical physics and carbide fuel development carried out 1978 for the Gas Cooled Fast Breeder Reactor at the Swiss Federal Institute for Reactor Research is described. (Auth.)

  20. Gas cooled fast reactor research and development program

    International Nuclear Information System (INIS)

    The research and development work in the field of core thermal-hydraulics, steam generator research and development, experimental and analytical physics and carbide fuel development carried out 1978 for the Gas Cooled Fast Breeder Reactor at the Swiss Federal Institute for Reactor Research is described. (Auth.)

  1. Efficient Water Management in Water Cooled Reactors

    International Nuclear Information System (INIS)

    number of the countries that have recently begun to consider the introduction of nuclear power are in water scarce regions, which would certainly limit the possibility for deployment of nuclear power plants, in turn hindering these countries' development and energy security. Thus, there is a large incentive to enhance efforts to introduce innovative water use, water management practices and related technologies. Water management for nuclear power plants is gaining interest in IAEA Member States as an issue of vital importance for the deployment of nuclear power. Recent experience has shown that some nuclear power plants are susceptible to prolonged drought conditions, forcing reactors to be shut down or power to be reduced to a minimal level. In some cases, environmental issues have resulted in regulations that limit the possibility for water withdrawal as well as water discharge. Regarding the most common design for cooling nuclear power plants, this has led to a complicated siting procedure for new plants and expensive retrofits for existing ones. The IAEA has already provided its Member States with reports and documents that address the issue. At the height of nuclear power expansion in the 1970s, the need for guidance in the area resulted in publications such as Thermal Discharges at Nuclear Power Stations - Their Management and Environmental Impact (Technical Reports Series No. 155) and Environmental Effects of Cooling Systems (Technical Reports Series No. 202). Today, amid the so-called nuclear renaissance, it is of vital importance to offer guidance to the Member States on the issues and possibilities that nuclear power water management brings. Management of water at nuclear power plants is an important subject during all phases of the construction, operation and maintenance of any nuclear power plant. Water management addresses the issue of securing water for condenser cooling during operation, for construction (during the flushing phase), and for inventory

  2. 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.

  3. Technology of steam generators for gas-cooled reactors. Proceedings of a specialists' meeting

    International Nuclear Information System (INIS)

    The activity of the IAEA in the field of the technology of gas-cooled reactors was formalized by formation of an International Working Group on Gas-Cooled Reactors (IWGCR). The gas cooled reactor program considered by the IWGCR includes carbon-dioxide-cooled thermal reactors, helium cooled thermal high temperature reactors for power generation and for process heat applications and gas-cooled fast breeder reactors. This report covers the papers dealing with operating experience, steam generators for next generation of gas-cooled reactors, material development and corrosion problems, and thermohydraulics

  4. 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.

  5. Pin-Type Gas Cooled Reactor for Nuclear Electric Propulsion

    Science.gov (United States)

    Wright, Steven A.; Lipinski, Ronald J.

    2003-01-01

    This paper describes a point design for a pin-type Gas-Cooled Reactor concept that uses a fuel pin design similar to the SP100 fuel pin. The Gas-Cooled Reactor is designed to operate at 100 kWe for 7 years plus have a reduced power mode of 20% power for a duration of 5 years. The power system uses a gas-cooled, UN-fueled, pin-type reactor to heat He/Xe gas that flows directly into a recuperated Brayton system to produce electricity. Heat is rejected to space via a thermal radiator that unfolds in space. The reactor contains approximately 154 kg of 93.15 % enriched UN in 313 fuel pins. The fuel is clad with rhenium-lined Nb-1Zr. The pressures vessel and ducting are cooled by the 900 K He/Xe gas inlet flow or by thermal radiation. This permits all pressure boundaries to be made of superalloy metals rather than refractory metals, which greatly reduces the cost and development schedule required by the project. The reactor contains sufficient rhenium (a neutron poison) to make the reactor subcritical under water immersion accidents without the use of internal shutdown rods. The mass of the reactor and reflectors is about 750 kg.

  6. Status of liquid metal cooled fast reactor technology

    International Nuclear Information System (INIS)

    During the period 1985-1998, there have been substantial advances in fast reactor technology development. Chief among these has been the demonstration of reliable operation by several prototypes and experimental reactors, the reliable operation of fuel at high burnup. At the IAEA meetings on liquid metal cooled fast reactor technology (LMFR), it became evident that there have been significant technological advances as well as changes in the economic and regulatory environment since 1985. Therefore the International working group on Fast Reactors has recommended the preparation of a new status report on fast reactors. The present report intends to provide comprehensive and detailed information on LMFR technology. The focus is on practical issues that are useful to engineers, scientists, managers, university students and professors, on the following topics: experience in construction and operation, reactor physics and safety, sore structural material and fuel technology, fast reactor engineering and activities in progress on LMFR plants

  7. Containment for Heavy-Water Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    The safety principles applicable to heavy-water, gas-cooled reactors are outlined, with a view to establishing containment specifications adapted to the sites available in Switzerland for the construction of nuclear plants. These specifications are derived from dose rates considered acceptable, in the event of a serious reactor accident, for persons living near the plant, and are based on-meteorological and demographic conditions representative of the majority of the country's sites. The authors consider various designs for the containment shell, taking into account the conditions which would exist in the shell after the maximum credible accident. The following types of shell are studied: pre-stressed concrete; pre-stressed concrete with steel dome; pre-stressed concrete with inner, leakproof steel lining; steel with concrete side shield to protect against radiation; double shell. The degree of leak proofing of the shells studied is regarded as a feature of the particular design and not as a fixed constructional specification. The authors assess the leak proofing properties of each type of shell and establish building costs for each of them on the basis of precise plans, with the collaboration of various specialized firms. They estimate the effectiveness of the various shells from a safety standpoint, in relation to different emergency procedures, in particular release into the atmosphere through appropriate filters and decontamination of the air within the shell by recycling through batteries of filters. The paper contains a very detailed comparison of about 10 cases corresponding to various combinations of design and emergency procedure; the comparison was made using a computer programme specially established for the purpose. The results are compared with those for a reactor of the same type and power, but assembled together with the heat exchangers in a pre-stressed concrete shell. (author)

  8. Transient Performance of Air-cooled Condensing Heat Exchanger in Long-term Passive Cooling System during Decay Heat Load

    International Nuclear Information System (INIS)

    In the event of a 'loss of coolant accident'(LOCA) and a non-LOCA, the secondary passive cooling system would be activated to cool the steam in a condensing heat exchanger that is immersed in an emergency cooldown tank (ECT). Currently, the capacities of these ECTs are designed to be sufficient to remove the sensible and residual heat from the reactor coolant system for 72 hours after the occurrence of an accident. After the operation of a conventional passive cooling system for an extended period, however, the water level falls as a result of the evaporation from the ECT, as steam is emitted from the open top of the tank. Therefore, the tank should be refilled regularly from an auxiliary water supply system when the system is used for more than 72 hours. Otherwise, the system would fail to dissipate heat from the condensing heat exchanger due to the loss of the cooling water. Ultimately, the functionality of the passive cooling system would be seriously compromised. As a passive means of overcoming the water depletion in the tank, Kim et al. applied for a Korean patent covering the concept of a long-term passive cooling system for an ECT even after 72 hours. This study presents transient performance of ECT with installing air-cooled condensing heat exchanger under decay heat load. The cooling capacity of an air-cooled condensing heat exchanger was evaluated to determine its practicality

  9. 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.

  10. Fuel Development For Gas-Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    M. K. Meyer

    2006-06-01

    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 [ ]. 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 a dispersion coated fuel particles in a ceramic matrix. Alternative fuel concepts considered in the U.S. 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.

  11. Design study on sodium-cooled large-scale reactor

    International Nuclear Information System (INIS)

    In Phase 1 of the 'Feasibility Study on Commercialized Fast Reactor Cycle Systems (F/S)', an advanced loop type reactor has been selected as a promising concept of sodium-cooled large-scale reactor, which has a possibility to fulfill the design requirements of the F/S. In Phase 2 of the F/S, it is planed to precede a preliminary conceptual design of a sodium-cooled large-scale reactor based on the design of the advanced loop type reactor. Through the design study, it is intended to construct such a plant concept that can show its attraction and competitiveness as a commercialized reactor. This report summarizes the results of the design study on the sodium-cooled large-scale reactor performed in JFY2001, which is the first year of Phase 2. In the JFY2001 design study, a plant concept has been constructed based on the design of the advanced loop type reactor, and fundamental specifications of main systems and components have been set. Furthermore, critical subjects related to safety, structural integrity, thermal hydraulics, operability, maintainability and economy have been examined and evaluated. As a result of this study, the plant concept of the sodium-cooled large-scale reactor has been constructed, which has a prospect to satisfy the economic goal (construction cost: less than 200,000yens/kWe, etc.) and has a prospect to solve the critical subjects. From now on, reflecting the results of elemental experiments, the preliminary conceptual design of this plant will be preceded toward the selection for narrowing down candidate concepts at the end of Phase 2. (author)

  12. Design requirements, operation and maintenance of gas-cooled reactors

    International Nuclear Information System (INIS)

    At the invitation of the Government of the USA the Technical Committee Meeting on Design Requirements, Operation and Maintenance of Gas-Cooled Reactors, was held in San Diego on September 21-23, 1988, in tandem with the GCRA Conference. Both meetings attracted a large contingent of foreign participants. Approximately 100 delegates from 18 different countries participated in the Technical Committee meeting. The meeting was divided into three sessions: Gas-cooled reactor user requirement (8 papers); Gas-cooled reactor improvements to facilitate operation and maintenance (10 papers) and Safety, environmental impacts and waste disposal (5 papers). A separate abstract was prepared for each of these 23 papers. Refs, figs and tabs

  13. Evaluation for External Reactor Vessel Cooling System using CFD Simulation

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Seok Bin; Park, Seong Dae; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2012-05-15

    To ensure the safety of the nuclear plants, there are lots of safety systems in the nuclear plant. One of them is External Reactor Vessel Cooling system (ERVC) which is operated when a molten corium is relocated in a lower head of a reactor vessel. As ERVC system runs, coolant flows down into a reactor cavity to remove a decay heat from the molten corium. This work simulated the ERVC system which is applied to APR1400 with CFD. To estimate the efficiency of the ERVC system, we designed the reactor cavity of the ERVC system of APR1400 in a full scale. From the designed model, we measured temperature distribution of the reactor vessel outer wall. Two kinds of coolant were used in this computational approach. One is present flooding matter which is water. The other is liquid metal gallium. With varying the area of the inlet and outlet of reactor cavity, we evaluated the importance of each variable

  14. Preliminary evaluation of alternate-fueled gas cooled fast reactors

    International Nuclear Information System (INIS)

    A preliminary evaluation of various alternative fuel cycles for the Gas-Cooled Fast Reactor (GCFR) is presented. Both homogeneous and heterogeneous oxide-fueled GCFRs are considered. The scenario considered is the energy center/dispersed reactor concept in which proliferation-resistant denatured reactors are coupled to 233U production reactors operating in secure energy centers. Individual reactor performance characteristics and symbiotic system parameters are summarized for several possible alternative fuel concepts. Comparisons are made between the classical homogeneous GCFR and the advanced heterogeneous concept on the basis of breeding ratio, doubling time, and net fissile gain. In addition, comparisons are made between a three-dimensional reactor model and the R-Z heterogeneous configuration utilized for the depletion and fuel management calculations. Lastly, thirty-year mass balance data are given for the various GCFR fuel cycles studied

  15. Gas-cooled fast breeder reactor. Quarterly progress report, February 1-April 30, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1980-05-01

    Information is presented concerning the reactor vessel; reactivity control mechanisms and instrumentation; reactor internals; primary coolant circuits;core auxiliary cooling system; reactor core; systems engineering; and reactor safety and reliability;

  16. 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.

  17. Feasibility and deployment strategy of water cooled thorium breeder reactor

    International Nuclear Information System (INIS)

    The author have studied water cooled thorium breeder reactor based on matured pressurized water reactor (PWR) plant technology for several years. Through these studies it is concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array with heavy water coolant in the primary loop by replacing original light water is appropriate for achieving sufficient breeding performance as sustainable fission system and high enough burn-up as an economical power plant. The heavy water cooled thorium reactor is feasible to be introduced by using Pu recovered from spent fuel of LWR, keeping continuity with current LWR infrastructure. This thorium reactor can be operated as sustainable energy supplier and also MA transmuter to realize future society with less long-lived nuclear waste

  18. Gas-cooled reactor safety and accident analysis

    International Nuclear Information System (INIS)

    The Specialists' Meeting on Gas-Cooled Reactor Safety and Accident Analysis was convened by the International Atomic Energy Agency in Oak Ridge on the invitation of the Department of Energy in Washington, USA. The meeting was hosted by the Oak Ridge National Laboratory. The purpose of the meeting was to provide an opportunity to compare and discuss results of safety and accident analysis of gas-cooled reactors under development, construction or in operation, to review their lay-out, design, and their operational performance, and to identify areas in which additional research and development are needed. The meeting emphasized the high safety margins of gas-cooled reactors and gave particular attention to the inherent safety features of small reactor units. The meeting was subdivided into four technical sessions: Safety and Related Experience with Operating Gas-Cooled Reactors (4 papers); Risk and Safety Analysis (11 papers); Accident Analysis (9 papers); Miscellaneous Related Topics (5 papers). A separate abstract was prepared for each of these papers

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

    International Nuclear Information System (INIS)

    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

  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. 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.

  2. Capital cost: gas cooled fast reactor plant

    International Nuclear Information System (INIS)

    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

  3. A study of occupant cooling by personally controlled air movement

    OpenAIRE

    Arens, Edward A.; Xu, T.; Miura, K; Zhang, H.; Fountain, M.; Bauman, Fred

    1997-01-01

    This study addresses the effectiveness of air movement cooling, an alternative to compressor-based cooling of the air itself. Subjects in an environmental chamber were exposed to a range of warm temperatures and allowed to adjust air movement to suit their individual preferences, while answering a series of questions about their comfort. Air movement was from the subject's side, in two modes of turbulent flow. The air speeds chosen by the subjects, and their subjective responses, are evaluate...

  4. 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.

  5. Gamma spectroscopy in water cooled reactors

    International Nuclear Information System (INIS)

    Gamma spectroscopy analysis of spent fuels in power reactors; study of two typical cases: determination of the power distribution by the mean of the activity of a low periodic element (Lanthanum 140) and determination of the burnup absolute rate by examining the ratio of Cesium 134 and Cesium 137 activities. Measures were realized on fuel solutions and on fuel assemblies. Development of a power distribution map of the assemblies and comparison with the results of a three dimensional calculation of core evolution

  6. 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.

  7. Consequences of reactor fuel damage: - Production of radioactive wastes. - Radioactivity in the reactor cooling system

    International Nuclear Information System (INIS)

    The report describes the consequences of damage of reactor fuel cladding. The types of damage and the release of fission products into the reactor cooling system are described as well as detection methods. The report also gives suggestions to reduce the consequences of a damage. (62 figs., 13 tabs.)

  8. Windscale advanced gas-cooled reactor (WAGR) decommissioning project overview

    International Nuclear Information System (INIS)

    The current BNFL reactor decommissioning projects are presented. The projects concern power reactor sites at Berkely, Trawsfynydd, Hunterstone, Bradwell, Hinkley Point; UKAEA Windscale Pile 1; Research reactors within UK Scottish Universities at East Kilbride and ICI (both complete); WAGR. The BNFL environmental role include contract management; effective dismantling strategy development; implementation and operation; sentencing, encapsulation and transportation of waste. In addition for the own sites it includes strategy development; baseline decommissioning planning; site management and regulator interface. The project objectives for the Windscale Advanced Gas-Cooled Reactor (WAGR) are 1) Safe and efficient decommissioning; 2) Building of good relationships with customer; 3) Completion of reactor decommissioning in 2005. The completed WAGR decommissioning campaigns are: Operational Waste; Hot Box; Loop Tubes; Neutron Shield; Graphite Core and Restrain System; Thermal Shield. The current campaign is Lower Structures and the remaining are: Pressure vessel and Insulation; Thermal Columns and Outer Vault Membrane. An overview of each campaign is presented

  9. New concept of proliferation resistant sodium cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Eliseev, V.A.; Krivitski, I.Y.; Matveev, V.I.; Popov, E.P.; Savitski, V.I.; Tsikunov, A.G. [Institute for Physics and Power Engineering, Obninsk (Russian Federation)

    2001-07-01

    The full text follows. It is proposed the concept of BN-800 sodium cooled fast reactor operating in the closed fuel cycle with special reprocessing technology. The use of nitride fuel allows improving the parameters of reactor safety (internal breeding {approx}1, zero value of sodium void reactivity effect), economy (one refueling per year), ecology (use of nitride enriched by nitrogen-15) and non-proliferation (use of reprocessing without separating the plutonium from uranium). The main difficulty of this type reactor development is that the technical project of BN-800 reactor with MOX fuel was developed. When using the nitride fuel it is necessary to serve (in max extent) the mail technical decisions of this project. This report presents first results on development and justification of the BN-800 reactor with nitride fuel core. (authors)

  10. IAEA activities in Gas-cooled Reactor technology development

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) has the charter to ''foster the exchange of scientific and technical information'', and ''encourage and assist research on, and development and practical application of, atomic energy for peaceful uses throughout the world''. This paper describes the Agency's activities in Gas-cooled Reactor (GCR) technology development

  11. IAEA activities in gas-cooled reactor technology development

    International Nuclear Information System (INIS)

    The International Atomic Energy Agency (IAEA) has the charter to ''foster the exchange of scientific and technical information'', and ''encourage and assist research on, and development and practical application of, atomic energy for peaceful uses throughout the world''. This paper describes the Agency's activities in Gas-cooled Reactor (GCR) technology development

  12. 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.

  13. Radiolytic reactions in the coolant of helium cooled reactors

    International Nuclear Information System (INIS)

    The success of helium cooled reactors is dependent upon the ability to prevent significant reaction between the coolant and the other components in the reactor primary circuit. Since the thermal reaction of graphite with oxidizing gases is rapid at temperatures of interest, the thermal reactions are limited primarily by the concentration of impurity gases in the helium coolant. On the other hand, the rates of radiolytic reactions in helium are shown to be independent of reactive gas concentration until that concentration reaches a very low level. Calculated steady-state concentrations of reactive species in the reactor coolant and core burnoff rates are presented for current U. S. designed, helium cooled reactors. Since precise base data are not currently available for radiolytic rates of some reactions and thermal reaction rate data are often variable, the accuracy of the predicted gas composition is being compared with the actual gas compositions measured during startup tests of the Fort Saint Vrain high temperature gas-cooled reactor. The current status of these confirmatory tests is discussed. 12 references

  14. Medium-size high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    This report summarizes high-temperature gas-cooled reactor (HTGR) experience for the 40-MW(e) Peach Bottom Nuclear Generating Station of Philadelphia Electric Company and the 330-MW(e) Fort St. Vrain Nuclear Generating Station of the Public Service Company of Colorado. Both reactors are graphite moderated and helium cooled, operating at approx. 7600C (14000F) and using the uranium/thorium fuel cycle. The plants have demonstrated the inherent safety characteristics, the low activation of components, and the high efficiency associated with the HTGR concept. This experience has been translated into the conceptual design of a medium-sized 1170-MW(t) HTGR for generation of 450 MW of electric power. The concept incorporates inherent HTGR safety characteristics [a multiply redundant prestressed concrete reactor vessel (PCRV), a graphite core, and an inert single-phase coolant] and engineered safety features

  15. Liquid-cooled nuclear reactor, especially a boiling water reactor

    International Nuclear Information System (INIS)

    A nuclear reactor with a special arrangement of fuel rods in the core is designed. Each fuel element has its shaft which is made of sheets, has the same cross section as the fuel element and protrudes at least the length of the control rod above the reactor core. Made of a zirconium alloy in the core area and of stainless steel above it, the shaft is equipped with channels for sliding the rods in and out and serves to spatially secure the position of the rods. Coolant flow is provided by the chimney effect. The shaft can conveniently enclose the control rod drive. It can also serve to bear the water separator. Moreover, it can constitute a part of the casing which surrounds the fuel rods and keeps the fuel in an intimate contact with the coolant; the other part of this casing is constituted by inserted sheets which can conveniently have the shape of angles. The walls of neighboring shafts form a compartment accommodating a neutron absorber plate. (M.D.). 11 figs

  16. 78 FR 64027 - Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors

    Science.gov (United States)

    2013-10-25

    ... COMMISSION Preoperational Testing of Emergency Core Cooling Systems for Pressurized-Water Reactors AGENCY... Core Cooling Systems for Pressurized-Water Reactors.'' This RG is being revised to incorporate guidance... emergency core cooling systems (ECCSs) of pressurized water reactors (PWRs). This RG also describes...

  17. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors AGENCY: Nuclear...-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling- Water Reactors.'' This... testing features of emergency core cooling systems (ECCSs) for boiling-water reactors (BWRs)....

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

    International Nuclear Information System (INIS)

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

  19. Cooling Performance of ALIP according to the Air or Sodium Cooling Type

    International Nuclear Information System (INIS)

    ALIP pumps the liquid sodium by Lorentz force produced by the interaction of induced current in the liquid metal and their associated magnetic field. Even though the efficiency of the ALIP is very low compared to conventional mechanical pumps, it is very useful due to the absence of moving parts, low noise and vibration level, simplicity of flow rate regulation and maintenance, and high temperature operation capability. Problems in utilization of ALIP concern a countermeasure for elevation of internal temperature of the coil due to joule heating and how to increase magnetic flux density of Na channel gap. The conventional ALIP usually used cooling methods by circulating the air or water. On the other hand, GE-Toshiba developed a double stator pump adopting the sodium-immersed self-cooled type, and it recovered the heat loss in sodium. Therefore, the station load factor of the plant could be reduced. In this study, the cooling performance with cooling types of ALIP is analyzed. We developed thermal analysis models to evaluate the cooling performance of air or sodium cooling type of ALIP. The cooling performance is analyzed for operating parameters and evaluated with cooling type. 1-D and 3-D thermal analysis model for IHTS ALIP was developed, and the cooling performance was analyzed for air or sodium cooling type. The cooling performance for air cooling type was better than sodium cooling type at higher air velocity than 0.2 m/s. Also, the air temperature of below 270 .deg. demonstrated the better cooling performance as compared to sodium

  20. 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

  1. Startup of the FFTF sodium cooled reactor. [Acceptance Test Program

    Energy Technology Data Exchange (ETDEWEB)

    Redekopp, R.D.; Umek, A.M.

    1981-03-01

    The Fast Flux Test Facility (FFTF), located on the Department of Energy (DOE) Hanford Reservation near Richland, Washington, is a 3 Loop 400 MW(t) sodium cooled fast reactor with a primary mission to test fuels and materials for development of the Liquid Metal Fast Breeder Reactor (LMFBR). Bringing FFTF to a condition to accomplish this mission is the goal of the Acceptance Test Program (ATP). This program was the mechanism for achieving startup of the FFTF. Highlights of the ATP involving the system inerting, liquid metal and inerted cell testing and initial ascent to full power are discussed.

  2. 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.

  3. 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.

  4. 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.

  5. Appraisal of possible combustion hazards associated with a high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    The report presents a study of combustion hazards that may be associated with the High Temperature Gas Cooled Reactor (HTGR) in the event of a primary coolant circuit depressurization followed by water or air ingress into the prestressed concrete reactor vessel (PCRV). Reactions between graphite and steam or air produce the combustible gases H2 and/or CO. When these gases are mixed with air in the containment vessel (CV), flammable mixtures may be formed. Various modes of combustion including diffusion or premixed flames and possibly detonation may be exhibited by these mixtures. These combustion processes may create high over-pressure, pressure waves, and very hot gases within the CV and hence may threaten the structural integrity of the CV or damage the instrumentation and control system installations within it. Possible circumstances leading to these hazards and the physical characteristics related to them are delineated and studied in the report

  6. Fuel performance and fission product behaviour in gas cooled reactors

    International Nuclear Information System (INIS)

    The Co-ordinated Research Programme (CRP) on Validation of Predictive Methods for Fuel and Fission Product Behaviour was organized within the frame of the International Working Group on Gas Cooled Reactors. This International Working Group serves as a forum for exchange of information on national programmes, provides advice to the IAEA on international co-operative activities in advanced technologies of gas cooled reactors (GCRs), and supports the conduct of these activities. The objectives of this CRP were to review and document the status of the experimental data base and of the predictive methods for GCR fuel performance and fission product behaviour; and to verify and validate methodologies for the prediction of fuel performance and fission product transport

  7. Decay heat removal in GEN IV gas cooled fast reactors

    International Nuclear Information System (INIS)

    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.

  8. Economic competitiveness requirements for evolutionary water cooled reactors

    International Nuclear Information System (INIS)

    This paper analyses the necessary economic conditions for evolutionary water cooled reactors to be competitive. Utilising recent national cost data for fossil-fired base load plants expected to be commissioned by 2005 -2010, target costs for nuclear power plants are discussed. Factors that could contribute to the achievement of those targets by evolutionary water cooled reactors are addressed. The feed-back from experience acquired in implementing nuclear programmes is illustrated by some examples from France and the Republic of Korea. The paper discusses the impacts on nuclear power competitiveness of globalisation and deregulation of the electricity market and privatisation of the electricity sector. In addition, issues related to external cost internalisation are considered. (author)

  9. Analysis of sodium-cooled fast reactor operations world-wide and consequences for future reactor design and operation

    International Nuclear Information System (INIS)

    In 2007, the sodium-cooled fast reactors which are currently operating or have operated throughout the world will have accumulated collective experience of 379 years of operation for 18 different reactors. This paper summarizes the various incidents and problems which have impacted these reactors' operations, ranks them by function and lists the solutions which were brought. This paper solely looks at the operating problems which occurred in the 'sodium' part of the fast reactors. Problems encountered in the classic water/steam part are not described herein review is made for all these reactors on: The water/sodium reactions occurred in the steam generators; -The technical difficulties on the primary components (pumps and exchangers) and the repair operations; -The incidents in handling operations; - The spurious leaks or transfers of sodium and their consequences; - The intakes of air or impurities; -The experience from fuel and clad failures; - The neutronic operations and control; -The material behaviour problems (as 321 or SPX1 drum); - The difficulties due to sodium aerosols. And all the specific difficulties due to sodium related technological problems. This paper also describes how the experience gained has been taken into account in the safe operation of these plants and also in the design of future reactors. n conclusion it appears that a significant experience has been accumulated that allow today good availability for the remaining operating plants and also good design possibilities for the future. (author)

  10. Air cooled turbine component having an internal filtration system

    Science.gov (United States)

    Beeck, Alexander R.

    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.

  11. A passive decay heat removal system for LWRs based on air cooling

    Energy Technology Data Exchange (ETDEWEB)

    Mochizuki, Hiroyasu, E-mail: mochizki@u-fukui.ac.jp [Research Institute of Nuclear Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan); Yano, Takahiro [Graduate School of Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan)

    2015-05-15

    Highlights: • A passive decay heat removal system for LWRs is discussed. • An air cooler model which condenses steam is developed. • The decay heat can be removed by air coolers with forced convection. • The dimensions of the air cooler are proposed. - Abstract: The present paper describes the capability of an air cooling system (ACS) to remove decay heat from a core of LWR such as an advanced boiling water reactor (ABWR) and a pressurized water reactor (PWR). The motivation of the present research is the Fukushima severe accident (SA) on 11 March 2011. Since emergency cooling systems using electricity were not available due to station blackout (SBO) and malfunctions, many engineers might understand that water cooling was not completely reliable. Therefore, a passive decay heat removal (DHR) system would be proposed in order to prevent such an SA under the conditions of an SBO event. The plant behaviors during the SBO are calculated using the system code NETFLOW++ for the ABWR and PWR with the ACS. Two types of air coolers (ACs) are applied for the ABWR, i.e., a steam condensing air cooler (SCAC) of which intake for heat transfer tubes is provided in the steam region, and single-phase type of which intake is provided in the water region. The DHR characteristics are calculated under the conditions of the forced air circulation and also the natural air convection. As a result of the calculations, the decay heat can be removed safely by the reasonably sized ACS when heat transfer tubes are cooled with the forced air circulation. The heat removal rate per one finned heat transfer tube is evaluated as a function of air flow rate. The heat removal rate increases as a function of the air flow rate.

  12. Advanced technologies for water cooled reactors 1990. Pt. 2

    International Nuclear Information System (INIS)

    The main purpose of the meeting was to review and discuss the status of national programmes, the progress achieved since the last meeting held in June 1988 in the field of advanced technologies and design trends for existing and future water cooled reactors. 24 specialists from 14 countries and the IAEA took part in the meeting and 12 papers were presented. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  13. The status of graphite development for gas cooled reactors

    International Nuclear Information System (INIS)

    The meeting was convened by the IAEA on the recommendation of the International Working Group on Gas Cooled Reactors. It was attended by 61 participants from 6 countries. The meeting covered the following subjects: overview of national programs; design criteria, fracture mechanisms and component test; materials development and properties; non-destructive examination, inspection and surveillance. The participants presented 33 papers on behalf of their countries. A separate abstract was prepared for each of these papers. Refs, figs, tabs, photos and diagrams

  14. Advanced technologies for water cooled reactors 1990. Pt. 1

    International Nuclear Information System (INIS)

    The meeting was attended by 20 participants from 12 countries who reviewed and discussed the status and progress of national programmes on advanced water-cooled reactors and recommended to the Scientific Secretary a comprehensive programme for 1991/1992 which would support technology development programmes in IWGATWR Member States. This summary report outlines the activities of IWGATWR since its Second Meeting in June 1988 and main results of the Third Meeting

  15. Safety Research Experiment Facility Project. Conceptual design report. Volume VII. Reactor cooling

    International Nuclear Information System (INIS)

    The Reactor Cooling System (RCS) will provide the required cooling during test operations of the Safety Research Experiment Facility (SAREF) reactor. The RCS transfers the reactor energy generated in the core to a closed-loop water storage system located completely inside the reactor containment building. After the reactor core has cooled to a safe level, the stored heat is rejected through intermediate heat exchangers to a common forced-draft evaporative cooling tower. The RCS is comprised of three independent cooling loops of which any two can remove sufficient heat from the core to prevent structural damage to the system components

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

    International Nuclear Information System (INIS)

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

  17. Description of the advanced gas cooled type of reactor (AGR)

    International Nuclear Information System (INIS)

    The present report comprises a technical description of the Advanced Gas cooled Reactor (AGR), a reactor type which has only been built in Great Britain. 14 AGR reactors have been built, located at 6 different sites and each station is supplied with twin-reactors. The Torness AGR 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 6 stations, Dungeness B, Hinkely Point B, Hunterston G, Hartlepool, Heysham I and Heysham II, are given only in tables with a summary of design data. Where specific data for Torness AGR has not been available, corresponding data from other AGR plans has been used, primarily from Heysham II, which belongs to the same generation of AGR reactors. The information presented is based on the open literature. The report is written as a part of the NKS/RAK-2 subproject 3: 'Reactors in Nordic Surroundings', which comprises a description of nuclear power plants neighbouring the Nordic countries. (au) 11 refs

  18. Stability monitoring of a natural-circulation-cooled boiling water reactor

    International Nuclear Information System (INIS)

    Methods for monitoring the stability of a boiling water reactor (BWR) are discussed. Surveillance of BWR stability is of importance as problems were encountered in several large reactors. Moreover, surveying stability allows plant owners to operate at high power with acceptable stability margins. The results of experiments performed on the Dodewaard BWR (the Netherlands) are reported. This type reactor is cooled by natural circulation, a cooling principle that is also being considered for new reactor designs. The stability of this reactor was studied both with deterministic methods and by noise analysis. Three types of stability are distinguished and were investigated separately: reactor-kinetic stability, thermal-hydraulic stability and total-plant stability. It is shown that the Dodewaard reactor has very large stability margins. A simple yet reliable stability criterion is introduced. It can be derived on-line from thhe noise signal of ex-vessel neutron detectors during normal operation. The sensitivity of neutron detectors to in-core flux perturbations - reflected in the field-of-view of the detector - was calculated in order to insure proper stability surveillance. A novel technique is presented which enables the determination of variations of the in-core coolant velocity by noise correlation. The velocity measured was interpreted on the basis of experiments performed on the air/water flow in a model of a BWR coolant channel. It appeared from this analysis that the velocity measured was much higher than the volume-averaged water and air velocities and the volumetric flux. The applicability of the above-mentioned technique to monitoring of local channel-flow stability was tested. It was observed that stability effects on the coolant velocity are masked by other effects originating from the local flow pattern. Experimental and theoretical studies show a shorter effective fuel time constant in a BWR than was assumed. (author). 118 refs.; 73 figs.; 21 tabs

  19. Thermophysical properties of materials for water cooled reactors

    International Nuclear Information System (INIS)

    The IAEA Co-ordinated Research Programme (CRP) to establish a thermophysical properties data base for light and heavy water reactor materials was organized within the framework of the IAEA's International Working Group on Advanced Technologies for Water Cooled Reactors. The work within the CRP started in 1990. The objective of the CRP was to collect and systemaize a thermophysical properties data base for light and heavy water reactor materials under normal operating, transient and accident conditions. The important thermophysical properties include thermal conductivity, thermal diffusivity, specific heat capacity, enthalpy, thermal expansion and others. These properties as well as the oxidation of zirconium-based alloys, the thermophysical characteristics of high temperature concrete-core melt interaction and the mechanical properties of construction materials are presented in this report. It is hoped that this report will serve as a useful source of thermophysical properties data for water cooled reactor analyses. The properties data are maintained on the THERSYST system at the University of Stuttgart, Germany and are internationally available. Refs, figs, tabs

  20. Modification of the Core Cooling System of TRIGA 2000 Reactor

    Science.gov (United States)

    Umar, Efrizon; Fiantini, Rosalina

    2010-06-01

    To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3-D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24°C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow's velocity in the inside of chimney.

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

    International Nuclear Information System (INIS)

    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 (CO2) 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)

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

    International Nuclear Information System (INIS)

    development or being considered for future water-cooled reactors; - Advantages that Gen III reactors have over previous designs in terms of economics, fuel utilisation, thermal efficiency, etc; - Operational issues of nuclear power plants in future low carbon energy systems with high shares of variable renewables, and issues posed by climate change (e.g. water scarcity, increased air and water temperatures and extreme weather events). - Standardisation, modularization and constructability issues and challenges; - A discussion of key differences between Gen II and Gen III designs, and possibilities of back-fitting Generation II reactors with new technologies, as part of a Long Term Operation strategy. This document brings together the available presentations (slides), dealing with: 1 - Utility safety and performance requirements in Europe (J. Spiler); 2 - EPRI Utility Requirement Document (S.B. Kim); 3 - WENRA activities on new and existing reactors (F. Feron); 4 - Evolution of the Finnish safety regulations and implementation ( M.L. Jaervinen); 5 - Multinational Design Evaluation Programme (J. Husse); 6 - EDF France modernization program for the existing NPPs (G. Ferraro); 7 - Innovations in GEN III designs and modernisation of existing NPP - An operator's point of view (F. Bertels); 8 - Modernisation of existing NPPs in Switzerland (W. Denk); 9 - Nuclear Technology Improvements in Modernization, Refurbishment and New Build Projects in Finland (H. Tuomisto); 10 - Round table discussion - Renewable and nuclear energy-based mitigation of climate change: substitution for fossil fuel usage (M. Golay); 11 - Innovation in water cooled reactor technologies (B. de l'Epinois); 12 - APR1400 - Safe, Reliable Technology (S. W. Kim); 13 - Advanced safety features of 3. generation VVER Plants (J. Laaksonen); 14 - Additional information on modern VVER GEN III Technology (M. Maltsev); 15 - Research and Development on Advanced PWR Design Improvement and Innovation in NPI (C. Yu); 16 - GE

  3. Liquid metal cooled reactors: Experience in design and operation

    International Nuclear Information System (INIS)

    on key fast reactor technology aspects in an integrative sense useful to engineers, scientists, managers, university students and professors. This publication has been prepared to contribute toward the IAEA activity to preserve the knowledge gained in the liquid metal cooled fast reactor (LMFR) technology development. This technology development and experience include aspects addressing not only experimental and demonstration reactors, but also all activities from reactor construction to decommissioning. This publication provides a survey of worldwide experience gained over the past five decades in LMFR development, design, operation and decommissioning, which has been accumulated through the IAEA programmes carried out within the framework of the TWG-FR and the Agency's INIS and NKMS

  4. Gas Reactor International Cooperative Program. Interim report. Construction and operating experience of selected European Gas-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    1978-09-01

    The construction and operating experience of selected European Gas-Cooled Reactors is summarized along with technical descriptions of the plants. Included in the report are the AVR Experimental Pebble Bed Reactor, the Dragon Reactor, AGR Reactors, and the Thorium High Temperature Reactor (THTR). The study demonstrates that the European experience has been favorable and forms a good foundation for the development of Advanced High Temperature Reactors.

  5. Storage of HLW in engineered structures: air-cooled and water-cooled concepts

    International Nuclear Information System (INIS)

    A comparative study on an air-cooled and a water-cooled intermediate storage of vitrified, highly radioactive waste (HLW) in overground installations has been performed by Nukem and Belgonucleaire respectively. In the air-cooled storage concept the decay heat from the storage area will be removed using natural convection. In the water-cooled storage concept the decay heat is carried off by a primary and secondary forced-cooling system with redundant and diverse devices. The safety study carried out by Nukem used a fault tree method. It shows that the reliability of the designed water-cooled system is very high and comparable to the inherent, safe, air-cooled system. The impact for both concepts on the environment is determined by the release route, but even during accident conditions the release is far below permissible limits. The economic analysis carried out by Belgonucleaire shows that the construction costs for both systems do not differ very much, but the operation and maintenance costs for the water-cooled facility are higher than for the air cooled facility. The result of the safety and economic analysis and the discussions with the members of the working group have shown some possible significant modifications for both systems, which are included in this report. The whole study has been carried out using certain national criteria which, in certain Member States at least, would lead to a higher standard of safety than can be justified on any social, political or economic grounds

  6. 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.

  7. 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.

  8. Physical and technical aspects of lead cooled fast reactors safety

    Energy Technology Data Exchange (ETDEWEB)

    Orlov, V.V.; Smirnov, V.S.; Filin, A.I. [Research and Development Institute of Power Engineering, Moscow (Russian Federation)

    2001-07-01

    The safety analysis of lead-cooled fast reactors has been performed for the well-developed concept of BREST-OD-300 reactor. The most severe accidents have been considered. An ultimate design-basis accident has been defined as an event resulting from an external impact and involving a loss of leak-tightness of the lead circuit, loss of forced circulation of lead and loss of heat sink to the secondary circuit, failure of controls and of reactor scram with resultant insertion of total reactivity margin, etc. It was assumed in accident analysis that the protective feature available for accident mitigation was only reactivity feedback on the changes in the temperatures of the reactor core elements and coolant flow rate, and in some cases also actuation of passive protections of threshold action in response to low flow rate and high coolant temperature at the core outlet. It should be noted that the majority of the analyzed accidents could be overcame even without initiation of the above protections. It has been demonstrated that a combination of inherent properties of lead coolant, nitride fuel, physical and design features of fast reactors will ensure natural safety of BREST and are instrumental for avoiding by a deterministic approach the accidents associated with a significant release of radioactivity and requiring evacuation of people in any credible initiating event and a combination of events. (author)

  9. The generation IV gas-cooled fast reactor

    International Nuclear Information System (INIS)

    The gas cooled fast reactor (GFR) is a helium-cooled fast spectrum reactor operating within a closed fuel cycle. It combines the advantages of fast reactors, in terms of a more sustainable use of uranium resources and waste minimisation, with the wider applicability of high temperature gas reactors, in terms of high efficiency electricity generation and the co-generation of high-quality process heat. Other advantages like the absence of threshold effect due to phase changing, the optical transparency and chemical inertness of the Helium coolant are also acknowledged. Within the European Union, GFR is one of the three fast reactors proposed for development to the demonstration stage within the European Sustainable Nuclear Industry Initiative (ESNII). On a wider global scale, GFR is one of the six systems proposed for further development within the Generation IV International Forum (GIF). In this respect, France, Switzerland, Japan and the European Union (through EURATOM) are signatories to the 'System Arrangement', the instrument through which the international research efforts are coordinated. This paper presents the current status of the development of the GFR system. The status of the GFR programme in each of the signatory countries is summarised including the intended contribution of the newly launched EURATOM 7. Framework Programme project - GoFastR. France has provided the bulk of the effort on conceptual design, safety assessment and fuel development. Switzerland makes significant contributions to the GFR system in the areas of core physics, uncertainty analysis, deterministic safety assessment and fuel development. Historically Japan has been very active in the development of the GFR system. Within the Generation IV GFR system, Japan contributes to the development of fuel and core materials

  10. 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.

  11. 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.

  12. 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.

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

    International Nuclear Information System (INIS)

    Graphical abstract: Display Omitted Research highlights: →20F and 23Ne tagging agents are produced by fast neutron flux. →20F 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. →20F and 23Ne power measurement has a reduced temperature influence. → Burn-up impact could be limited by simultaneous 20F and 23Ne 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,α) 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.

  14. Specific power of liquid-metal-cooled reactors

    International Nuclear Information System (INIS)

    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

  15. 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

  16. 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",

  17. 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.

  18. Study on a decay heat removal system of light water reactors using air coolers

    International Nuclear Information System (INIS)

    In the present work, a passive decay heat removal system for light water reactors (LWRs) based on a new concept is studied referring to an air cooling system (ACS) of the fast breeder reactor Monju. The present study will contribute to the reduction of severe accident risks of nuclear power plants. In this system, a blower for an air cooler (AC) is operated using the rotation of a small steam turbine by generated steam in order to cool heat transfer tubes by forced convection of air. The purpose of the present work is to investigate the plant transient caused by a station blackout (SBO) using the plant system code NETFLOW++ and decay heat removal characteristics. A calculation model is the Advanced Boiling Water Reactor (ABWR) in Japan. (author)

  19. Gas Cooled Fast Reactors: Recent advances and prospects

    International Nuclear Information System (INIS)

    The paper presents the current status of the Gas cooled Fast Reactor system development which is shared within the Generation IV International Forum including EURATOM through the 7th Framework Programme project GoFastR. The various areas considered will include suitable fuel compounds and high temperature resistant cladding materials options, core design optimisation, primary system boundary, energy conversion. The safety approach, mainly oriented on core cooling for the moment, will be recalled together with a discussion of the results obtained. Further potential improvements or simplification of the system safety, at the light of the Fukushima accident, including an indirect coupled cycle for the energy conversion and a self sustainable Decay Heat Removal loop will be mentioned. The main issues related to the necessary R&D programme accompanying the system development will be recalled (fuel and materials, helium coolant technology, components such as gas circulators, valves and heat exchangers, thermal barriers). (author)

  20. Gas cooled fast reactor background, facilities, industries and programmes

    International Nuclear Information System (INIS)

    This report was prepared at the request of the OECD-NEA Coordinating Group on Gas Cooled Fast Reactor Development and it represents a contribution (Vol.II) to the jointly sponsored Vol.I (GCFR Status Report). After a chapter on background with a brief description of the early studies and the activities in the various countries involved in the collaborative programme (Austria, Belgium, France, Germany, Japan, Sweden, Switzerland, United Kingdom and United States), the report describes the facilities available in those countries and at the Gas Breeder Reactor Association and the industrial capabilities relevant to the GCFR. Finally the programmes are described briefly with programme charts, conclusions and recommendations are given. (orig.)

  1. Cooling molten salt reactors using “gas-lift”

    Energy Technology Data Exchange (ETDEWEB)

    Zitek, Pavel, E-mail: zitek@kke.zcu.cz, E-mail: klimko@kke.zcu.cz; Valenta, Vaclav, E-mail: zitek@kke.zcu.cz, E-mail: klimko@kke.zcu.cz; Klimko, Marek, E-mail: zitek@kke.zcu.cz, E-mail: klimko@kke.zcu.cz [University of West Bohemia in Pilsen, Univerzitní 8, 306 14 Pilsen (Czech Republic)

    2014-08-06

    This study briefly describes the selection of a type of two-phase flow, suitable for intensifying the natural flow of nuclear reactors with liquid fuel - cooling mixture molten salts and the description of a “Two-phase flow demonstrator” (TFD) used for experimental study of the “gas-lift” system and its influence on the support of natural convection. The measuring device and the application of the TDF device is described. The work serves as a model system for “gas-lift” (replacing the classic pump in the primary circuit) for high temperature MSR planned for hydrogen production. An experimental facility was proposed on the basis of which is currently being built an experimental loop containing the generator, separator bubbles and necessary accessories. This loop will model the removal of gaseous fission products and tritium. The cleaning of the fuel mixture of fluoride salts eliminates problems from Xenon poisoning in classical reactors.

  2. Cooling molten salt reactors using "gas-lift"

    Science.gov (United States)

    Zitek, Pavel; Valenta, Vaclav; Klimko, Marek

    2014-08-01

    This study briefly describes the selection of a type of two-phase flow, suitable for intensifying the natural flow of nuclear reactors with liquid fuel - cooling mixture molten salts and the description of a "Two-phase flow demonstrator" (TFD) used for experimental study of the "gas-lift" system and its influence on the support of natural convection. The measuring device and the application of the TDF device is described. The work serves as a model system for "gas-lift" (replacing the classic pump in the primary circuit) for high temperature MSR planned for hydrogen production. An experimental facility was proposed on the basis of which is currently being built an experimental loop containing the generator, separator bubbles and necessary accessories. This loop will model the removal of gaseous fission products and tritium. The cleaning of the fuel mixture of fluoride salts eliminates problems from Xenon poisoning in classical reactors.

  3. Gas-cooled reactor coolant circulator and blower technology

    International Nuclear Information System (INIS)

    In the previous 17 meetings held within the framework of the International Working Group on Gas-Cooled Reactors, a wide variety of topics and components have been addressed, but the San Diego meeting represented the first time that a group of specialists had been convened to discuss circulator and blower related technology. A total of 20 specialists from 6 countries attended the meeting in which 15 technical papers were presented in 5 sessions: circulator operating experience I and II (6 papers); circulator design considerations I and II (6 papers); bearing technology (3 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

  4. Thorium fueled high temperature gas cooled reactors. An assessment

    International Nuclear Information System (INIS)

    The use of thorium as a fertile fuel for the High Temperature Gas Cooled Reactor (HTR) instead of uranium has been reviewed. It has been concluded that the use of thorium might be beneficial to reduce the actinide waste production. To achieve a real advancement, the uranium of the spent fuel has to be recycled and the requested make-up fissile material for the fresh fuel has to be used in the form of highly-enriched uranium. A self-sustaining fuel cycle may be possible in the HTR of large core size, but this could reduce the inherent safety features of the design. (orig.)

  5. SIMMER-III modeling of gas cooled fast reactor

    International Nuclear Information System (INIS)

    This paper deals with extension and application of the SIMMER-III code for safety studies of a gas cooled fast reactor. The equation of state of the helium gas and its thermal physical properties have been prepared and implemented in the code. The geometric, thermal hydraulic and neutronic models have been set up for the ALLEGERO reactor. The code and the associated model are verified by comparing steady state and unprotected loss of flow 20% remained flow rate (ULOF-20%) results with those done by other project partners. Reasonable or good agreements have been achieved for major physical variables. The unprotected loss of coolant accident (ULOCA) case is a severe transient case with core melting and degradation that was emulated only by SIMMER, in the project. In the initiating phase the clad becomes molten, this triggers the first power excursion. Then the fuel becomes more mobile and further power excursions take place, which lead to core melting and degradation. The fuel is ejected by power excursion and then moves relatively slowly to the lower part of vessel. Finally there are only a few kilograms of fuel escaping to the vessel outside (into reactor container) and the released thermal energy is about 6 GJ within a period of one minute. The final power stays below one MW and the reactor is in a deep sub-criticality state, since 1/2 fuel becomes noneffective. (author)

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

    International Nuclear Information System (INIS)

    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 MWth 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

  7. 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.

  8. 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)

  9. Varying duty operation of air-cooled condenser units

    Science.gov (United States)

    Milman, O. O.; Kondratev, A. V.; Ptakhin, A. V.; Dunaev, S. N.; Kirjukhin, A. V.

    2016-05-01

    Results of experimental investigations of operation modes of air-cooled condensers (ACC) under design and varying duty conditions are presented. ACCs with varying cooling airflow rates under constant heat load and with constant cooling airflow under varying heat load are examined. Diagrams of heat transfer coefficients and condensation pressures on the heat load and cooling airflow are obtained. It is found that, if the relative heat load is in the range from 0.6 to 1.0 of the nominal value, the ACC heat transfer coefficient varies insignificantly, unlike that of the water-cooled surface condensers. The results of the determination of "zero points" are given, i.e., the attainable pressure in air-cooled condensing units (ACCU), if there is no heat load for several values of working water temperature at the input of water-jet ejectors and liquid ring vacuum pump. The results of the experimental determination of atmospheric air suction into the ACC vacuum system. The effect of additional air suctions in the steam pipe on ACCU characteristics is analyzed. The thermal mapping of ACC heat exchange surfaces from the cooling air inlet is carried out. The dependence of the inefficient heat exchange zone on the additional air suction into the ACC vacuum system is given. It is shown that, if there is no additional air suction into the ACC vacuum system, the inefficient heat exchange zone is not located at the bottom of the first pass tubes, and their portion adjacent to the bottom steam pipe works efficiently. Design procedures for the ACC varying duty of capacitors are presented, and their adequacy for the ACCU varying duty estimation is analyzed.

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

    OpenAIRE

    Mr. Bhikhu B,; Prof. Ronak Shah

    2014-01-01

    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 ...

  11. 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.

  12. Optimum Reactor Outlet Temperatures for High Temperature Gas-Cooled Reactors Integrated with Industrial Processes

    Energy Technology Data Exchange (ETDEWEB)

    Lee O. Nelson

    2011-04-01

    This report summarizes the results of a temperature sensitivity study conducted to identify the optimum reactor operating temperatures for producing the heat and hydrogen required for industrial processes associated with the proposed new high temperature gas-cooled reactor. This study assumed that primary steam outputs of the reactor were delivered at 17 MPa and 540°C and the helium coolant was delivered at 7 MPa at 625–925°C. The secondary outputs of were electricity and hydrogen. For the power generation analysis, it was assumed that the power cycle efficiency was 66% of the maximum theoretical efficiency of the Carnot thermodynamic cycle. Hydrogen was generated via the hightemperature steam electrolysis or the steam methane reforming process. The study indicates that optimum or a range of reactor outlet temperatures could be identified to further refine the process evaluations that were developed for high temperature gas-cooled reactor-integrated production of synthetic transportation fuels, ammonia, and ammonia derivatives, oil from unconventional sources, and substitute natural gas from coal.

  13. BREST-OD-300 Reactor as a prototype of the future commercial lead cooled fast reactor of natural safety

    Energy Technology Data Exchange (ETDEWEB)

    Orlov, V.V.; Smirnov, V.S.; Filin, A.I.; Glazov, A.G. [N.A. Dollezhal Institute ' NIKIET' , PO Box 788, Moscow, 101000 (Russian Federation)

    2006-07-01

    This paper briefly describes the physical and design features of a demonstration 300 MWe fast reactor with uranium-plutonium nitride fuel and lead coolant, BREST-OD-300, under development in Russia. This reactor is regarded as a prototype of future commercial reactors, which may form a foundation for large-scale growth of nuclear power in this new century. It is demonstrated that the natural properties of the lead coolant and nitride fuel combined with the physical and design features specific to fast reactors ensure natural safety of BREST and, with any credible initiating events, allow deterministic exclusion of accidents with large radioactive releases requiring evacuation of local residents. The paper identifies the ways and means of attaining natural safety, which rule out prompt criticality excursion, loss of cooling and fuel failure through use of a small reactivity margin, commensurable with {beta}{sub eff}, low pressure in the circuit, large margins to temperature limits, high natural circulation, passive decay heat removal by air unlimited in time, high heat accumulating capability of lead-filled circuit, stabilizing temperature and coolant flow rate feedbacks, etc. (authors)

  14. Bruce NGS A/B assessment of reactor vault fans on air mixing patterns

    International Nuclear Information System (INIS)

    The development and results of numerical simulations of air mixing patterns in the CANDU Bruce Nuclear Generating Station reactor vault, as a function of vault cooling fan combinations, are presented. The results of this analysis will serve as a basis for selecting appropriate fan combination scenarios to consider in the upcoming post-LOCA (loss of coolant) hydrogen-air-steam mixing analysis. Following a severe reactor accident in which fuel cooling is impaired, a significant amount of hydrogen may be produced from the steam/Zircaloy reaction and subsequently released into containment. The hydrogen ignition system mitigates the consequences of hydrogen burns to within acceptable safety limits. Igniters deliberately initiate a burn of the hydrogen-air-steam mixture as it reaches its flammability limits. However without adequate mixing, the igniters may become blinded by a region of non-flammable hydrogen mixture while an unfavourable hydrogen mixture forms elsewhere. The vault cooling fans play an important role in promoting mixing in the vault atmosphere. To help assess the effects of vault cooling fans on air mixing, an analysis was carried out to identify the air mixing patterns as a function of different fan availability combinations. The three-dimensional containment code, GOTHIC, was used to model the Bruce containment with modelling emphasis on the reactor vault geometry and the vault cooling system fans. Twenty-five fan combination air mixing simulations and eight tracer gas fan dispersion simulations were performed. The results showed that air mixing patterns created by individual fans can be superimposed to determine the effects of various fan combinations, there was symmetry of flow patterns between the west and east vault halves, and there was a general absence of significant stagnant regions in the reactor vault. 8 refs., 7 figs., 3 tabs

  15. ELECTRA: A European Lead-bismuth Cooled Training Reactor

    International Nuclear Information System (INIS)

    The lack of low power liquid metal cooled reactors has meant that few engineers within the nuclear power industry and research community are familiar with operational procedures of this family of coolants, expected to be used for Generation IV fast neutron systems. The reasons for this lack may include safety issues related to use of Mercury, NaK or sodium as applied in early low power reactors. Especially in western Europe, no low power liquid metal cooled reactor was ever in operation. Here, we present the design of a 2 MWth lead-bismuth cooled reactor with (Pu,Zr)N fuel, relying on natural convection for full power operation. The combination of low power density with natural convection for heat removal makes the reactor ideal for training purposes. The large thermal expansion of heavy liquid metals makes is possible to design low power fast neutron reactors relying on natural convection. Since at present, there exists no suitable material for pumps operating at high velocity in lead alloy environments, the only possible short term solution for constructing such a reactor is anyway to design for 100% natural circulation of the coolant. For this purpose, a small core height, a large difference between coolant inlet and outlet temperatures and a low coolant velocity is desired. By application of (Pu,Zr)N fuel, criticality can be achieved with a fissile inventory of 100 kg LWR grade plutonium. Monte Carlo simulations show that 19 hexagonal fuel assemblies, each with 91 fuel pins having an outer diameter of 1.1 cm, and an active height of 15 cm is sufficient to obtain a critical core. Including end pellets, gas plenum and end caps, the total core height is limited to 30 cm. Adopting P/D = 1.25 and a heat exchanger elevation of 4 m, it is found that 2 MW of thermal power may be removed by a natural circulation velocity of 0.4 m/s. This corresponds to a linear rating of 8 kW/m and a temperature increase of the coolant equal to 240 degrees. Limiting the clad temperature

  16. 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.

  17. 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.

  18. Incorporation of statistical distribution of particle properties in chemical reactor design and operation: the cooled tubular reactor

    OpenAIRE

    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 theory presented is illustrated with a number of examples. It is shown that pellet-scale statistics have an impact on cooled tubular reactor design and operation. Cooled tubular reactor design is d...

  19. Passive Safety Optimization in Liquid Sodium-Cooled Reactors

    International Nuclear Information System (INIS)

    The governing equations and the solutions schemes are developed for the three-dimensional thermal-hydraulic model.A detailed constitutive relations are also developed through the analysis with a CFD code. The developed model is able to obtain a detailed thermal hydraulic information in a subassemlby of a liquid metal-cooled reactor core. The model has been integrated with the system analysis codes SASSYS-1 and SSC-K to be validated for the SHRT-17 test performed in the EBR-II reactor. The baseline analyses were performed with the customary thermal-hydraulic model and with the new model for the reference design of KALIMER-150. The results point out that increased detail in the thermal-hydraulic analysis must be matched by increased detail in the reactivity feedback modeling, especially the radial core expansion model, in order to provide distinguishable differences. The analyses performed to identify certain key safety features have indicated that the adjustment of core restraint system is able to provide beneficial bending of hexcans, resulting in enhanced negative reactivity feedback in unprotected accidents. Tasks were performed to identify and assess the implications for plant safety of proposed specific approaches for reducing capital and operating costs of next generation sodium-cooled fast reactors. A new plant design of the sodium-cooled reactor concept is developed utilizing a gas turbine Brayton cycle, which uses supercritical carbon dioxide (S-CO2) as the working fluid. In addition, the design innovations are incorporated in modular sodium-to-supercritical carbon dioxide heat exchangers that enable the traditional intermediate heat transport circuit to be eliminated. Several evaluations and analyses for the safety design and efficiency suggest that the developed system is safe and cost-effective. Test plans are developed for the measurement of phenomenological data describing freezing of molten metallic fuel, melt relocation and interaction with steel

  20. Present Status of Canadian Organic-Cooled Reactor Technology

    International Nuclear Information System (INIS)

    Canada has been operating the first organic-cooled D2O-moderated pressure-tube reactor (WR-1) since November 1965. The operation of WR-1 and the supporting development programmes at the Whiteshell Nuclear Research Establishment are of direct interest to those seeking to exploit the advantages of organic coolants. The following contributions have been made which bear on the technical feasibility of the concept. 1. HB-40, a partially hydrogenated terphenyl mixture, has been used routinely in WR-1 at temperatures up to 370°C. It is now developed to the point where it is the major contender for use in a power reactor. It has the advantages of being liquid at room temperature and offering cost savings over other organic coolants at comparable conditions. 2. Zirconium base alloys can be made compatible with organic coolants provided 50 to 300 ppm of water are added to condition the surface of the zirconium alloy against hydrogen penetration. Chlorine contamination must be controlled. 3. Fuel can be operated at practical ratings without limitations imposed by fouling provided reasonable attention is paid to coolant chemistry. 4. Non-bonded uranium carbide fuels have been taken to bumups as high 11 000 MWd/t and would probably be acceptable at burnups in excess of 16 000 MWd/tU. These and other subjects associated with the present status of the technology are the subject of this paper. The excellent performance of WR-1 indicates that, as yet, no insurmountable technical limitation faces the designers of organic cooled reactors. (author)

  1. Hybrid high temperature gas-cooled reactor, thermonuclear fusion

    International Nuclear Information System (INIS)

    The project of a multi-purpose high temperature gas-cooled reactor started in 1969. The Atomic Energy Commission, Japan, approved in 1980 the budget for the design study of the experimental reactor. The conceptual design is in progress. The manufacturing of coated fuel pellets and the test method have been developed. The study of graphite structure is carried out. Corrosion and creep tests are made to obtain the knowledge concerning the metals in high temperature helium gas. The engineering study of various machines and structures operating at high temperature is performed. International cooperative works are considered. The experimental reactor will be critical in 1987. A critical plasma test facility, JT-60, has been constructed at the Japan Atomic Energy Research Institute. As the theoretical work on plasma confinement, the evaluation of the critical beta value of JT-60 was made. By high temperature neutral beam injection, the slowing down and heating processes of high energy particles are studied. The development of a non-circular cross-section tokamak is in progress. The construction of JT-60 will be completed in 1984. Study concerning superconducting magnets is considered. Japan is one of the members of INTOR project. (Kato, T.)

  2. Romanian Contribution to the Development of Lead Cooled Fast Reactors

    International Nuclear Information System (INIS)

    In Romania the nuclear energy is considered an important component of energy mix and for a country sustainable development. Presently based on PHWR CANDU technology, the research and development activities, part of the national nuclear power programme, provide an increased effort towards generation 4, dedicated to support fast reactor lead technology. In the European framework (EU R&D Framework Programmes, European Sustainable Nuclear Industry Initiative) devoted to the development of GenerationIV technologies, Romania is contributing as a partner in EU R&D projects together with a large number of EU research organizations and in the ALFRED MoU, having ANSALDO Nucleare, ENEA and INR as initial members. ALFRED (Advanced Lead Fast Reactor European Demonstrator) project aims to build a 125MWe lead cooled fast reactor demonstrator, connected to the electrical network, with a target date for operation start-up in 2025. In February 2011 Romanian Government approved the option to host ALFRED demonstrator. Based on the access to European structural funds, existing nuclear experience and EU orientation to build the demonstrators in the new member states, Romania is an important option for siting process. An investigation of the existing national capabilities, identification of additional infrastructure and identification of the professional development needs in order to prepare the siting national support are presented in the paper. The main approaches and needed resources to meet expected requirements for ALFRED implementation are discussed as well. (author)

  3. Overview of PbBi-Cooled Reactor Development and ADS Program in China

    International Nuclear Information System (INIS)

    Most of China's electricity is produced from fossil fuels. Rapid growth in demand has given rise to power shortages, and the reliance on fossil fuels has led to much air pollution. Nuclear power has an important role for the China energy safety and environmental protection. According to the current layout of nuclear power development, by around 2040, pressurized water reactors are expected to level off at 200 GWe. However, the nuclear fuel shortage and the spent fuel accumulation will restrict the healthy development of nuclear energy in China. So the innovative nuclear system is expected to solve these problems. Accelerator driven sub-critical system (ADS) is an important approach to incinerate the long-lived high-level nuclear waste, which is focused by many countries. Liquid Lead-Bismuth eutectic (LBE) has many unique nuclear, thermophysical and chemical attributes that is potential candidate coolant for ADS subcritical reactors. In addition, lead and bismuth can produce copious spallation neutrons when bombarded with energetic protons. This makes LBE one of the top candidates for a high-power spallation target and coolant in an accelerator-driven subcritical system. In China, the fundamental research of ADS was carried out 10 years ago supported by Ministry of Science and Technology. From 2009, Chinese Academy of Sciences (CAS) carried out a project of the ADS prophase research to develop the key technologies about superconducting accelerator, LBE loop, materials et al. In 2011, a large scaled ADS development program to transmute the nuclear waste has been launched by CAS named ''advanced nuclear fission energy''. In this program, CAS plan to develop the ADS system by 3 phases. In the first phase, form 2011-2017, an ADS verification facility named ADS-VF will be build. The ADS-VF consists of a LBE cooled reactor coupled with a proton accelerator and liquid metal spallation target. For the second phase in this program, an ADS experimental reactor named ADS

  4. 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

  5. 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.

  6. Generation IV nuclear energy system initiative; air-cooled option RCCS studies and NSTF preparation

    International Nuclear Information System (INIS)

    The work documented in this report is a follow-on to the FY 2010 work reported on the air-cooled RCCS design of the General Atomics (GA) Modular High Temperature Gas-cooled Reactor (MHTGR). This workscope focused on confirming the scaling laws selected to design the scaled experiments in the large-scale Argonne National Laboratory (ANL) Natural Convection Shutdown Heat Removal Test Facility (NSTF) and on sensitivity studies to evaluate the effect of the modeling approximations made in FY 2010 on the models developed for the experiment design supporting analysis. In addition design support analyses were performed to evaluate NSTF configuration modifications proposed to enhance instrumentation performance. The objectives were to: (a) update and confirm the scaling relations to be used in modifying the existing NSTF facility into a scaled model of the GA-MHTGR air-cooled RCCS; and (b) provide an analytical basis for the NSTF modifications needed to conduct scaled experimental simulations of this RCCS. In summary, the scaling evaluation updated the basis that the air-cooled RCCS can be simulated at the ANL NSTF facility at a prototypic scale in the lateral direction and about half scale in the vertical direction. The density behavior of air was modified. A parallel track approach was taken with both RELAP5 system evaluations and STAR-CCM+ computational fluid dynamic (CFD) evaluations carried out as the set of instrumentation configurations were detailed.

  7. 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.

  8. Development of failure detection system for gas-cooled reactor

    International Nuclear Information System (INIS)

    This work presents several kinds of Failure Detection Systems for Fuel Elements, stressing their functional principles and major applications. A comparative study indicates that the method of electrostatic precipitation of the fission gases Kr and Xe is the most efficient for fuel failure detection in gas-cooled reactors. A detailed study of the physical phenomena involved in electrostatic precipitation led to the derivation of an equation for the measured counting rate. The emission of fission products from the fuel and the ion recombination inside the chamber are evaluated. A computer program, developed to simulate the complete operation of the system, relates the counting rate to the concentration of Kr and Xe isotopes. The project of a mock-up is then presented. Finally, the program calculations are compared to experimental data, available from the literature, yielding a close agreement. (author)

  9. Advances in High Temperature Gas Cooled Reactor Fuel Technology

    International Nuclear Information System (INIS)

    This publication reports on the results of a coordinated research project on advances in high temperature gas cooled reactor (HTGR) fuel technology and describes the findings of research activities on coated particle developments. These comprise two specific benchmark exercises with the application of HTGR fuel performance and fission product release codes, which helped compare the quality and validity of the computer models against experimental data. The project participants also examined techniques for fuel characterization and advanced quality assessment/quality control. The key exercise included a round-robin experimental study on the measurements of fuel kernel and particle coating properties of recent Korean, South African and US coated particle productions applying the respective qualification measures of each participating Member State. The summary report documents the results and conclusions achieved by the project and underlines the added value to contemporary knowledge on HTGR fuel.

  10. 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.

  11. 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

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

    International Nuclear Information System (INIS)

    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

  13. 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).

  14. EVALUATION OF COOLING INSTRUMENTATION SYSTEM OF TRIGA MARK II REACTOR OF BANDUNG

    International Nuclear Information System (INIS)

    Evaluation of cooling instrumentation system of Triga Mark II reactor has been done. The reactor has been upgraded from 1 MW to 2 MW. The increasing of power is performed by changing the reactor components and systems. The reactor cooling system has important role in reactor operation, the system transfers heat produced in the core. The operation of the cooling system needed to be back up with qualified instrumentation. Evaluation has been done by doing analysis and observing the equipment design, type and clarification, performance study of instrumentation and system related to cooling system. It is known that the performance and system of Triga mark II reactor included the cooling system. It is also obtained the characteristic data of primary and secondary cooling system, piping diagram and instrumentation, emergency core cooling system. The cooling system has 4 measurement, i.e. flow rate, input and output temperature to heat exchanger, and electricity conductivity of water. The measurement can be observed from the reactor console. From this evaluation it is concluded that cooling system instrumentation followed the required criteria

  15. Chemistry control strategies for a supercritical water-cooled reactor

    International Nuclear Information System (INIS)

    The long-term viability of any Generation IV Supercritical Water-cooled Reactor (SCWR) concept depends on the ability of reactor designers and operators to predict and control water chemistry to minimize corrosion and corrosion product transport. Currently, SCWR material testing is being carried out using a limited range of water chemistries to establish the dependencies of the corrosion behavior on parameters such as water temperature and dissolved oxygen concentration. Once a final suite of candidate alloys is identified, more detailed, longer term testing will be needed under water chemistry regimes expected to be used in the SCWR. Prior to these tests, it will be necessary to identify proposed water chemistry regimes for the SCWR, and provide expected ranges for the key parameters. The direct-cycle design proposed for various SCWR concepts take advantage of the extensive operating experience world-wide of fossil-fired SCW power plants. Conceptually, the SCWR replaces the fossil-fired boiler with the reactor core (pressure vessel or pressure tube); the concept is broadly similar to that of a boiling water reactor. Current fossil-fired SCW power plants use either an all-volatile treatment or oxygenated water treatment for the feedwater systems. While the optimal water chemistry for a SCWR is yet to be determined, the monitored parameters are likely to be the same as those in existing fossil-fired and nuclear power plants: pH; conductivity, and concentrations of O2, H2, additives, impurities, corrosion and activation products. Monitoring will be required at many of the same components: feedwater, main 'steam', drains, pump outlets, condenser hotwell, and purification inlets and outlets. This paper outlines the strategy being used to develop a water chemistry regime for a CANDU® SCWR. It describes the key areas identified for chemistry monitoring and control: a) the reactor core, where materials are subjected to irradiation and high temperature, b

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-09-01

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

  17. Unsteady thermal analysis of gas-cooled fast reactor core

    International Nuclear Information System (INIS)

    This thesis presents numerical analysis of transient heat transfer in an equivalent coolant-fuel rod cell of a typical gas cooled, fast nuclear reactor core. The transient performance is assumed to follow a complete sudden loss of coolant starting from steady state operation. Steady state conditions are obtained from solving a conduction problem in the fuel rod and a parabolic turbutent convection problem in the coolant section. The coupling between the two problems is accomplished by ensuring continuity of the thermal conditions at the interface between the fuel rod and the coolant. to model turbulence, the mixing tenght theory is used. Various fuel rod configurations have been tested for optimal transient performance. Actually, the loss of coolant accident occurs gradually at an exponential rate. Moreover, a time delay before shutting down the reactor by insertion of control rods usually exists. It is required to minimize maximum steady state cladding temperature so that the time required to reach its limiting value during transient state is maximum. This will prevent the escape of radioactive gases that endanger the environment and the public. However, the case considered here is a limiting case representing what could actually happen in the worst probable accident. So, the resutls in this thesis are very indicative regarding selection of the fuel rode configuration for better transient performance in case of accidents in which complete loss of collant occurs instantaneously

  18. Thermohydraulics of emergency core cooling in light water reactors

    International Nuclear Information System (INIS)

    This report, by a group of experts of the OECD-NEA Committee on the Safety of Nuclear Installations, reviews the current state-of-knowledge in the field of emergency core cooling (ECC) for design-basis, loss-of-coolant accidents (LOCA) and core uncover transients in pressurized- and boiling-water reactors. An overview of the LOCA scenarios and ECC phenomenology is provided for each type of reactor, together with a brief description of their ECC systems. Separate-effects and integral-test facilities, which contribute to understanding and assessing the phenomenology, are reviewed together with similarity and scaling compromises. All relevant LOCA phenomena are then brought together in the form of tables. Each phenomenon is weighted in terms of its importance to the course of a LOCA, and appraised for the adequacy of its data base and analytical modelling. This qualitative procedure focusses attention on the modelling requirements of dominant LOCA phenomena and the current capabilities of the two-fluid models in two-phase flows. This leads into the key issue with ECC: quantitative code assessment and the application of system codes to predict with a well defined uncertainty the behaviour of a nuclear power plant. This issue, the methodologies being developed for code assessment and the question of how good is good enough are discussed in detail. Some general conclusions and recommendations for future research activities are provided

  19. Uncertainties in analysis of innovative lead-cooled fast reactors

    International Nuclear Information System (INIS)

    There are numerous uncertainties in the prediction of innovative reactor design, arising from approximations used in the solution of the transport equation, and in nuclear data processing and cross section libraries generation. This paper describes the problems encountered in the analysis of the Encapsulated Nuclear Heat Source (ENHS) core benchmark and the new cross section libraries developed to overcome these problems. The ENHS is a new lead-bismuth or lead cooled novel reactor concept that is fuelled with metallic alloy of Pu, U and Zr, and i designed to operate for 20 effective full power years without refuelling and with very small burnup reactivity swing. The computational tools benchmarked include MOCUP-a coupled MCNP-4C and ORIGEN2.1 utility codes with MCNP data libraries based on ENDF/B-VI evaluations; and KWO2-a coupled KENO-V.a and ORIGEN2.1 code with ENDFB-V.2 based 238 groups library. Uncertainties in the cross sections of lead were found particularly large and deserve careful evaluation. (author)

  20. Gas-Cooled Fast Reactor (GFR) FY04 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    K. D. Weaver; T. C. Totemeier; D. E. Clark; E. E. Feldman; E. A. Hoffman; R. B. Vilim; T. Y. C. Wei; J. Gan; M. K. Meyer; W. F. Gale; M. J. Driscoll; M. Golay; G. Apostolakis; K. Czerwinski

    2004-09-01

    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.

  1. Numerical simulation of sodium pool fires in liquid metal-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    In Liquid Metal-Cooled Fast Breeder Reactor (LMFBR), the leakage of sodium can result in sodium fires. Due to sodium's high chemical reactivity in contact with air and water, sodium fires will lead to an immediate increase of the air temperature and pressure in the containment. This will harm the integrity of the containment. In order to estimate and foresee the sequence of this accident, or to prevent the accident and alleviate the influence of the accident, it is necessary to develop programs to analyze such sodium fire accidents. Based on the work of predecessors, flame sheet model is produced and used to analyze sodium pool fire accidents. Combustion model and heat transfer model are included and expatiated. And the comparison between the analytical and experimental results shows the program is creditable and reasonable. This program is more realistic to simulate the sodium pool fire accidents and can be used for nuclear safety judgement. (authors)

  2. 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.

  3. 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.

  4. Safety aspects of the Modular High-Temperature Gas-Cooled Reactor (MHTGR)

    International Nuclear Information System (INIS)

    The Modular High-Temperature Gas-Cooled Reactor (MHTGR) is an advanced reactor concept under development through a cooperative program involving the US Government, the nuclear industry and the utilities. The design utilizes the basic high-temperature gas-cooled reactor (HTGR) features of ceramic fuel, helium coolant, and a graphite moderator. The qualitative top-level safety requirement is that the plant's operation not disturb the normal day-to-day activities of the public. The MHTGR safety response to events challenging the functions relied on to retain radionuclides within the coated fuel particles has been evaluated. A broad range of challenges to core heat removal have been examined which include a loss of helium pressure and a simultaneous loss of forced cooling of the core. The challenges to control of heat generation have considered not only the failure to insert the reactivity control systems, but the withdrawal of control rods. Finally, challenges to control chemical attack of the ceramic coated fuel have been considered, including catastrophic failure of the steam generator allowing water ingress or of the pressure vessels allowing air ingress. The plant's response to these extreme challenges is not dependent on operator action and the events considered encompass conceivable operator errors. In the same vein, reliance on radionuclide retention within the full particle and on passive features to perform a few key functions to maintain the fuel within acceptable conditions also reduced susceptibility to external events, site-specific events, and to acts of sabotage and terrorism. 4 refs., 14 figs., 1 tab

  5. High Temperature Gas Cooled Reactor Fuels and Materials

    International Nuclear Information System (INIS)

    At the third annual meeting of the technical working group on Nuclear Fuel Cycle Options and Spent Fuel Management (TWG-NFCO), held in Vienna, in 2004, it was suggested 'to develop manuals/handbooks and best practice documents for use in training and education in coated particle fuel technology' in the IAEA's Programme for the year 2006-2007. In the context of supporting interested Member States, the activity to develop a handbook for use in the 'education and training' of a new generation of scientists and engineers on coated particle fuel technology was undertaken. To make aware of the role of nuclear science education and training in all Member States to enhance their capacity to develop innovative technologies for sustainable nuclear energy is of paramount importance to the IAEA Significant efforts are underway in several Member States to develop high temperature gas cooled reactors (HTGR) based on either pebble bed or prismatic designs. All these reactors are primarily fuelled by TRISO (tri iso-structural) coated particles. The aim however is to build future nuclear fuel cycles in concert with the aim of the Generation IV International Forum and includes nuclear reactor applications for process heat, hydrogen production and electricity generation. Moreover, developmental work is ongoing and focuses on the burning of weapon-grade plutonium including civil plutonium and other transuranic elements using the 'deep-burn concept' or 'inert matrix fuels', especially in HTGR systems in the form of coated particle fuels. The document will serve as the primary resource materials for 'education and training' in the area of advanced fuels forming the building blocks for future development in the interested Member States. This document broadly covers several aspects of coated particle fuel technology, namely: manufacture of coated particles, compacts and elements; design-basis; quality assurance/quality control and characterization techniques; fuel irradiations; fuel

  6. Calculations on heavy-water moderated and cooled natural uranium fuelled power reactors

    International Nuclear Information System (INIS)

    One of the codes that the Instituto Nacional de Investigaciones Nucleares (Mexico) has for the nuclear reactors design calculations is the LEOPARD code. This work studies the reliability of this code in reactors design calculations which component materials are the same of the heavy water moderated and cooled, natural uranium fuelled power reactors. (author)

  7. Advanced water-cooled reactor technologies. Rationale, state of progress and outlook

    International Nuclear Information System (INIS)

    Eighty per cent of the world's power reactors are water cooled and moderated. Many improvements in their design and operation have been implemented since the first such reactor started commercial operation in 1957. This report addresses the safety, environmental and economic rationales for further improvements, as well as their relevance to currently operating water reactors

  8. 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.

  9. Reactor fault simulation at the closure of the Windscale advanced gas-cooled reactor: analysis of reactor transient tests

    International Nuclear Information System (INIS)

    The testing of fault transient analysis methods by direct simulation of fault sequences on a commercial reactor is clearly excluded on safety and economic grounds. The closure of the Windscale prototype advanced gas-cooled reactor (WAGR) therefore offered a unique opportunity to test fault study methods under extreme conditions relatively unfettered by economic constraints, although subject to appropriate safety regulations. One aspect of these important experiments was a series of reactor transient tests. The objective of these reactor transients was to increase confidence in the fault study computer models used for commercial AGR safety assessment by extending their range of validation to cover large amplitude and fast transients in temperature, power and flow, relevant to CAGR faults, and well beyond the conditions achievable experimentally on commercial reactors. A large number of tests have now been simulated with the fault study code KINAGRAX. Agreement with measurement is very good and sensitivity studies show that such discrepancies as exist may be due largely to input data errors. It is concluded that KINAGRAX is able to predict steady state conditions and transient amplitudes in both power and temperature to within a few percent. (author)

  10. 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.

  11. 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)

  12. 10 CFR Appendix J to Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors

    Science.gov (United States)

    2010-01-01

    ... CFR 50.12, are still applicable to Option B of this appendix if necessary, unless specifically revoked...-Cooled Power Reactors J Appendix J to Part 50 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES Pt. 50, App. J Appendix J to Part 50—Primary Reactor...

  13. 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... Cooling Systems for New Boiling-Water Reactors.'' This RG describes testing methods the NRC staff...)-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors.''...

  14. Air Cooling for High Temperature Power Electronics (Presentation)

    Energy Technology Data Exchange (ETDEWEB)

    Waye, S.; Musselman, M.; King, C.

    2014-09-01

    Current emphasis on developing high-temperature power electronics, including wide-bandgap materials such as silicon carbide and gallium nitride, increases the opportunity for a completely air-cooled inverter at higher powers. This removes the liquid cooling system for the inverter, saving weight and volume on the liquid-to-air heat exchanger, coolant lines, pumps, and coolant, replacing them with just a fan and air supply ducting. We investigate the potential for an air-cooled heat exchanger from a component and systems-level approach to meet specific power and power density targets. A proposed baseline air-cooled heat exchanger design that does not meet those targets was optimized using a parametric computational fluid dynamics analysis, examining the effects of heat exchanger geometry and device location, fixing the device heat dissipation and maximum junction temperature. The CFD results were extrapolated to a full inverter, including casing, capacitor, bus bar, gate driver, and control board component weights and volumes. Surrogate ducting was tested to understand the pressure drop and subsequent system parasitic load. Geometries that met targets with acceptable loads on the system were down-selected for experimentation. Nine baseline configuration modules dissipated the target heat dissipation, but fell below specific power and power density targets. Six optimized configuration modules dissipated the target heat load, exceeding the specific power and power density targets. By maintaining the same 175 degrees C maximum junction temperature, an optimized heat exchanger design and higher device heat fluxes allowed a reduction in the number of modules required, increasing specific power and power density while still maintaining the inverter power.

  15. Unlimited cooling capacity of the passive-type emergency core cooling system of the MARS reactor

    International Nuclear Information System (INIS)

    The MARS nuclear plant is a 600 MWth PWR with completely passive core safeguards. The most relevant innovative safety system is the Emergency Core Cooling System (ECCS), which is based on natural circulation, and on a passive-type activation that follows a core flow decrease, whatever was the cause (only one component, 400% redundant, is not static). The main thermal hydraulic transients occurring as a consequence of design basis accidents for the MARS plant were presented at the ICONE 3 Conference. Those transients were analyzed in the first stage, with the aim at pointing out the capability of the innovative ECCS to intervene. So, they included only a short-time analysis (extended for a few hundreds of seconds) and the well known RELAP 5 computer program was used for this purpose. In the present paper, the long-term analyses (extended for several thousands of seconds) of the same transients are shown. These analyses confirmed that the performance of the Emergency Core Cooling System of the MARS reactor is guaranteed also in long-term scenarios

  16. Air quality impact analysis in support of the new production reactor environmental impact statement

    International Nuclear Information System (INIS)

    The Pacific Northwest Laboratory (PNL) conducted this air quality impact analysis for the US Department of Energy (DOE). The purpose of this work was to provide Argonne National Laboratory (ANL) with the required estimates of ground-level concentrations of five criteria air pollutants at the Hanford Site boundary from each of the stationary sources associated with the new production reactor (NPR) and its supporting facilities. The DOE proposes to provide new production capacity for the primary production of tritium and secondary production of plutonium to support the US nuclear weapons program. Three alternative reactor technologies are being considered by DOE: the light-water reactor, the low-temperature, heavy-water reactor, and the modular high-temperature, gas-cooled reactor. In this study, PNL provided estimates of the impacts of the proposed action on the ground-level concentration of the criteria air pollutants for each of the alternative technologies. The criteria pollutants were sulfur dioxide, nitrogen dioxide, carbon monoxide, total suspended particulates, and particulates with a diameter of less than 10 microns. Ground-level concentrations were estimated for the peak construction phase activities expected to occur in 1997 and for the operational phase activities beginning in the year 2000. Ground-level concentrations of the primary air pollutants were estimated to be well below any of the applicable national or state ambient air quality standards. 12 refs., 19 tabs

  17. 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

  18. 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

  19. Hydrogen in water-cooled nuclear power reactors

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  1. 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.

  2. Design measures in evolutionary water cooled reactors to optimize for economic viability

    International Nuclear Information System (INIS)

    Since the mid 1980s, there have been various efforts to develop evolutionary water cooled reactors based on the current operating plant experience. To sustain and improve the economic viability, particular attention has been paid to the following aspects in developing evolutionary water cooled reactors: design simplification and increased operating margins, standardization in design as well as construction and operation, integration of operating plant insights, and consideration of safety, operability and constructability during the design stage. This paper reviews each item and discusses several examples from some of the evolutionary water cooled reactors being developed. (author)

  3. Thermohydraulic relationships for advanced water cooled reactors, and the role of IAEA

    International Nuclear Information System (INIS)

    Under the auspices of the International Atomic Energy Agency (IAEA) a Coordinated Research Program (CRP) on Thermohydraulic Relationships for Advanced Water-Cooled Reactors was carried out from 1995-1999. It was included into the IAEA's Programme following endorsement in 1995 by the IAEA's International Working Group on Advanced Technologies for Water Cooled Reactors. The overall goal was to promote international information exchange and cooperation in establishing a consistent set of thermohydraulic relationships that are appropriate for use in analyzing the performance and safety of advanced water cooled reactors. (authors)

  4. 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.

  5. Water chemistry management in cooling system of research reactor in JAERI

    International Nuclear Information System (INIS)

    The department of research reactor presently operates three research reactors (JRR-2, JRR-3M and JRR-4). For controlling and management of water and gas in each research reactor are performed by the staffs of the research reactor technology development division. Water chemistry management of each research reactor is one of the important subject. The main objects are to prevent the corrosion of water cooling system and fuel elements, to suppress the plant radiation build-up and to minimize the radioactive waste. In this report describe a outline of each research reactor facilities, radiochemical analytical methods and chemical analytical methods for water chemistry management. (author)

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

    International Nuclear Information System (INIS)

    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

  7. 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.

  8. Evaluating Fault Detection and Diagnostics Protocols Applied to Air-Cooled Vapor Compression Air-Conditioners

    OpenAIRE

    Yuill, David P.; Braun, James E.

    2012-01-01

    Fault detection and diagnostics (FDD) tools are being increasingly applied in air-conditioning systems. There are many different protocols used in these FDD tools, so an important question to ask is: how well do the protocols work? This paper describes the ongoing development of the first standardized method of evaluation for FDD protocols applied to air-cooled vapor compression air-conditioning systems. The general approach is to feed a library of data – including temperatures, pressures, an...

  9. Effect mechanism of air deflectors on the cooling performance of dry cooling tower with vertical delta radiators under crosswind

    International Nuclear Information System (INIS)

    Highlights: • A 3D numerical model was set for NDDCTV to study the effect of air deflectors. • The air deflectors improve the tower performance by 1.375 °C at uc = 6 m/s for a case. • The air deflectors reduce the air inflow deviation angle θd at most delta entries. • The reduced θd can improve the cooling performance of former deteriorated columns. • Both the radial inflow air velocity and θd impact the cooling performance of delta. - Abstract: To study the effect mechanism of air deflectors on dry cooling tower, a three dimensional numerical model was established, with full consideration of the delta structure. The accuracy and credibility of dry cooling tower numerical model were validated. By numerical model, the average air static pressure and the average radial inflow air velocity were computed and analyzed at delta air entry, sector air entry and exit faces. By the air inflow deviation angle θd, the effect of air deflectors on the aerodynamic field around tower was analyzed. The water exit temperatures of θ−1 columns, θ+2 columns and cooling sectors were also presented to clarify the effect of air deflectors. It was found that the air deflectors improved the aerodynamic field around cooling columns. The reduced air inflow deviation degree at delta entry improved the cooling performance of deteriorated columns. Referring to the radial inflow air velocity ura and the air inflow deviation degree at delta entry, the effect mechanism of air deflectors are clarified under crosswind

  10. 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

  11. Decay heat removal and heat transfer under normal and accident conditions in gas cooled reactors

    International Nuclear Information System (INIS)

    The meeting was convened by the International Atomic Energy Agency on the recommendation of the IAEA's International Working Group on Gas Cooled Reactors. It was attended by participants from China, France, Germany, Japan, Poland, the Russian Federation, Switzerland, the United Kingdom and the United States of America. The meeting was chaired by Prof. Dr. K. Kugeler and Prof. Dr. E. Hicken, Directors of the Institute for Safety Research Technology of the KFA Research Center, and covered the following: Design and licensing requirements for gas cooled reactors; concepts for decay heat removal in modern gas cooled reactors; analytical methods for predictions of thermal response, accuracy of predictions; experimental data for validation of predictive methods - operational experience from gas cooled reactors and experimental data from test facilities. Refs, figs and tabs

  12. High temperature corrosion of structural materials under gas-cooled reactor helium

    International Nuclear Information System (INIS)

    The Generation IV International Forum has selected six promising nuclear power systems for further collaborative investigations and development. Among these six concepts, two candidates are Gas Cooled Reactors (GCR), namely the Very High Temperature Reactor (VHTR) and the Gas-cooled Fast Reactor (GFR). The CEA has launched a R and D program on the metallic materials for application in an innovative GCR. Structural GCR alloys have been extensively studied in the past three decades. Some critical aspects for the steels and nickel base alloys resistance under the service conditions are microstructural stability, creep strength and compatibility with the cooling gas. The coolant, namely helium, proved to contain impurities mainly H2, CO, CH4, N2 and steam in the microbar range that interact with metals at high temperature. Surface scale formation, bulk carburisation and/or decarburisation can occur, depending on the atmosphere characteristics, primarily the effective oxygen partial pressure and carbon activity, on the temperature and on the alloys chemical composition. These structural transformations can notably influence the mechanical properties: carburisation may induce a loss in toughness and ductility whereas decarburisation impedes the creep strength. There is a valuable theoretical as well as practical knowledge on the corrosion of high temperature alloys in the primary circuit of a GCR but this past experience is not sufficient to qualify every component in a future reactor. On the one hand, the material environment could be significantly different from the former GCR's, especially regarding the higher temperature. On the other hand, the materials of interest are partly different. Ni-Cr-W alloys, for instance, may offer significant improvement in the maximum operating temperature as far as the mechanical properties are concerned. However, their corrosion resistance toward the GCR atmosphere is still unknown. We describe here our first corrosion tests of Haynes

  13. Safety aspects of forced flow cooldown transients in Modular High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    During some of the design basis accidents in Modular High Temperature Gas Cooled Reactors (MHTGRs), the main Heat Transport System (HTS) and the Shutdown Cooling System n removed by the passive Reactor (SCS) are assumed to have failed. Decay heat is the Cavity Cooling System (RCCS) only. If either forced flow cooling system becomes available during such a transient, its restart could significantly reduce the down-time. This report used the THATCH code to examine whether such restart, during a period of elevated core temperatures, can be accomplished within safe limits for fuel and metal component temperatures. If the reactor is scrammed, either system can apparently be restarted at any time, without exceeding any safe limits. However, under unscrammed conditions a restart of forced cooling can lead to recriticality, with fuel and metal temperatures significantly exceeding the safety limits

  14. Safety aspects of forced flow cooldown transients in modular high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kroeger, P.G.

    1992-01-01

    During some of the design basis accidents in Modular High Temperature Gas Cooled Reactors (MHTGRs) the main Heat Transport System (HTS) and the Shutdown Cooling System (SCS), are assumed to have failed. Decay heat is then removed by the passive Reactor Cavity Cooling System (RCCS) only. If either forced flow cooling system becomes available during such a transient, its restart could significantly reduce the down-time. This paper uses the THATCH code to examine whether such restart, during a period of elevated core temperatures, can be accomplished within safe limits for fuel and metal component temperatures. If the reactor is scrammed, either system can apparently be restarted at any time, without exceeding any safe limits. However, under unscrammed conditions a restart of forced cooling can lead to recriticality, with fuel and metal temperatures significantly exceeding the safety limits.

  15. Safety aspects of forced flow cooldown transients in modular high temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kroeger, P.G.

    1992-09-01

    During some of the design basis accidents in Modular High Temperature Gas Cooled Reactors (MHTGRs) the main Heat Transport System (HTS) and the Shutdown Cooling System (SCS), are assumed to have failed. Decay heat is then removed by the passive Reactor Cavity Cooling System (RCCS) only. If either forced flow cooling system becomes available during such a transient, its restart could significantly reduce the down-time. This paper uses the THATCH code to examine whether such restart, during a period of elevated core temperatures, can be accomplished within safe limits for fuel and metal component temperatures. If the reactor is scrammed, either system can apparently be restarted at any time, without exceeding any safe limits. However, under unscrammed conditions a restart of forced cooling can lead to recriticality, with fuel and metal temperatures significantly exceeding the safety limits.

  16. Specialists' meeting on fission product release and transport in gas-cooled reactors. Summary report

    International Nuclear Information System (INIS)

    The purpose of the Meeting on Fission Product Release and Transport in Gas-Cooled Reactors was to compare and discuss experimental and theoretical results of fission product behaviour in gas-cooled reactors under normal and accidental conditions and to give direction for future development. The technical part of the meeting covered operational experience and laboratory research, activity release, and behaviour of released activity

  17. 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.

  18. 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.

  19. 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

  20. 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

  1. Research and development for high temperature gas cooled reactor in Japan

    International Nuclear Information System (INIS)

    The paper describes the current status of High Temperature Gas Cooled Reactor research and development work in Japan, with emphasis on the Experimental Very High Temperature Reactor (Exp. VHTR) to be built by Japan Atomic Energy Research Institute (JAERI) before the end of 1985. The necessity of construction of Exp. VHTR was explained from the points of Japanese energy problems and resources

  2. Performance Prediction Method of CO2 Cycle for Air Cooling

    Science.gov (United States)

    Koyama, Shigeru; Xue, Jun; Kuwahara, Ken

    From the perspective of global environmental protection and energy-saving, the research and development on high-efficiency heat pump and refrigeration systems using environment-friendly refrigerants have become one of the most important issues in the air-conditioning and refrigeration sector. In the present work, a steady-state model of the CO2 transcritical cycle for air cooling, which consists of a rotary compressor, a fin-tube gas cooler,a fin-tube evaporator and an expansion valve, has been developed. The detailed model of fin-tube heat exchanger has been constructed by means of the finite volume method, in which the local heat transfer and flow characteristics are evaluated. It should be noted that the effects of the dew condensation generated on the cooling surface are considered in the evaporator model. As a calculation example, the effects of the indoor air wet-bulb temperature on the cycle performance have been examined with this developed simulator.

  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. 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

  5. Cooling system for a prestressed concrete vessel of a nuclear reactor

    International Nuclear Information System (INIS)

    Cooling system for a prestressed concrete vessel of a nuclear reactor which is faced with a steel liner and an insulating layer, and provided with cooling tubes flown through by a coolant and fixed at the prestressed concrete vessel. The cooling tubes are placed side by side and combined into several, independently operable units; they are supplied with the coolant by distributing lines operating independently, and adjacent tubes are flown through by the flow medium in opposite direction. (orig.)

  6. 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.

  7. 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.

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

    International Nuclear Information System (INIS)

    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.

  9. 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.)

  10. Experimental studies on heat transfer in external cooling of the reactor pressure vessel

    International Nuclear Information System (INIS)

    The filling of the reactor cavity by accidental initiation of the containment spray system, while reactor is in full power, could have severe consequences. If the relatively cold water suddenly cools down the wall of fully pressurized reactor pressure vessel and a crack is assumed to be located on the outer surface of the vessel, the induced thermal stresses might damage the pressure vessel wall. The effects of the inadvertent cooling and pressurized thermal shock (PTS) were studied experimentally at Lappeenranta University of Technology and the heat transfer coefficients gained from the experimental results were compared with calculations. (author)

  11. Determination of the Design Speed of the Primary Cooling Pump in the Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Hyungi; Seo, Kyoungwoo; Chi, Daeyoung; Park, Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    An open-pool type research reactor is widely designed in consideration of the reactor operation and accessibility. Reactor structure assembly is generally placed at the pool bottom. rimary cooling system circulates the coolant from the reactor core to the heat exchanger. Therefore the heat generated from the reactor core is continuously removed. After the primary cooling pumps stop, the decay heat is removed by the coastdown flow induced by the inertia force of a flywheel attached to each primary cooling pump. A pump coastdown flow means that the pump operates with the angular momentums of the shaft, impeller, and flywheel when a loss of electricity occurs. The primary cooling pump consists of the pump, flywheel, and moto. They are connected by flexible couplings. The primary cooling pump is conceptually designed based on the required flow rate and system constraints. A centrifugal pump of Case 1 with a non-dimensional specific speed of 0.59 and specific diameter of 4.94 is chosen as the primary cooling pump based on the hydraulic performance and mechanical integrity.

  12. 78 FR 64029 - Cost-Benefit Analysis for Radwaste Systems for Light-Water-Cooled Nuclear Power Reactors

    Science.gov (United States)

    2013-10-25

    ... analysis for liquid and gaseous radwaste system components for light water nuclear power reactors... 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...

  13. Emergency cooling down of fast-neutron reactors by natural convection (a review)

    Science.gov (United States)

    Zhukov, A. V.; Sorokin, A. P.; Kuzina, Yu. A.

    2013-05-01

    Various methods for emergency cooling down of fast-neutron reactors by natural convection are discussed. The effectiveness of using natural convection for these purposes is demonstrated. The operating principles of different passive decay heat removal systems intended for cooling down a reactor are explained. Experimental investigations carried out in Russia for substantiating the removal of heat in cooling down fast-neutron reactors are described. These investigations include experimental works on studying thermal hydraulics in small-scale simulation facilities containing the characteristic components of a reactor (reactor core elements, above-core structure, immersed and intermediate heat exchangers, pumps, etc.). It is pointed out that a system that uses leaks of coolant between fuel assemblies holds promise for fast-neutron reactor cooldown purposes. Foreign investigations on this problem area are considered with making special emphasis on the RAMONA and NEPTUN water models. A conclusion is drawn about the possibility of using natural convection as the main method for passively removing heat in cooling down fast-neutron reactors, which is confirmed experimentally both in Russia and abroad.

  14. 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.

  15. Enhancing VHTR passive safety and economy with thermal radiation based direct reactor auxiliary cooling system

    International Nuclear Information System (INIS)

    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 RVACS can be characterized as a surface-based decay heat removal system. It 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 the core volume) and decay heat removal capability (proportional to the vessel surface area). 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 environmental side. For the reactor side, cooling pipes will be inserted into holes made in the outer or inner graphite reflector blocks. There will be gaps or annular regions formed between these cooling pipes and their corresponding surrounding graphite surfaces. Graphite has an excellent heat conduction property. By taking advantage of this feature, we can have a volume-based method to remove decay heat. The scalability can be achieved, if needed, by employing more rows of cooling pipes to accommodate higher decay heat rates. Since heat can easily conduct through the graphite regions among the holes made for the cooling pipes, those cooling pipes located further away from the active core region can still be very

  16. Numerical study by large-eddy simulation on effects and mechanism of air-cooling enhancing technologies

    International Nuclear Information System (INIS)

    Learning from the lessons of the Fukushima Daiichi nuclear incident in which a long-term station black-out occurred, we have been developing an air-cooling system for boiling water reactors that can operate without electricity for a virtually indefinite time. Improvement in the heat transfer performance of air-cooling is key to the development of the air-cooling system. We developed air-cooling enhancing technologies for the air-cooling system by using heat transfer fins, turbulence-enhancing ribs and a micro-fabrication surface. In our previous study, the performance of these air-cooling enhancing technologies was evaluated by heat transfer tests using a single pipe of the air-cooling heat exchanger. To achieve further improvement of the heat transfer performance, it is important to understand the mechanism of the air-cooling enhancing technologies. In this study, we used the numerical analysis which is based on the filtered incompressible Navier-Stokes equation and the filtered energy equation with the large-eddy simulation in order to investigate the effects and the mechanism of the developed air-cooling enhancing technologies. We found that the analysis results agreed well with the experimental results and the empirical formula results. The heat transfer enhancement mechanism of the heat transfer fin is due to an increase in the heat transfer area. Due to a decrease in the flow velocity at the base of the fins, the increase in the Nusselt number was approximately 15% smaller than the estimated value from the area increase. In the heat transfer enhancement by the turbulence-enhancing ribs, the unsteady behavior of the large-scale vortex generated by the flow separation plays an important role. The enhancement ratio of the Nusselt number by the micro-fabrication surface can be explained by the apparent thermal conductivity. The Nusselt number was increased 4-8% by the micro-fabrication surface. The effect of the micro-fabrication surface is increased by applying

  17. An experimental investigation of the air entrainment in the shutdown cooling system during mid-loop operation

    International Nuclear Information System (INIS)

    An experimental study on the air entrainment phenomena during mid-loop operation has been performed for Ulchin 3 and 4 nuclear power plant (UCN 3 and 4). The UCN 3 and 4 is the standard Combustion Engineering (CE) System 80, two-loop, 2825 MWt pressurized water reactor which is currently under construction in Korea. This study was undertaken by Korea Atomic Energy Research Institute to provide a basis for modification of CE system 80 design which has relatively small mid-loop operating range, and to investigate the impact of the air entrainment on the shutdown cooling pump. A 1/4 scale model test was performed for two shutdown cooling suction nozzle configurations, with and without bellmouth entry, to obtain data relative to air ingestion and vortex formation at the shutdown cooling suction nozzle during mid-loop operation. The test model size was determined based on Froude number corresponding to that which occurs in plants by considering that the Weber number and Reynolds number are large enough to ensure that liquid surface tension and viscosity would not significantly affect the vortex formation. An empirical correlation between the flow rate and the critical submergence was obtained with respect to the Froude number. The impact of the air entrainment of the pump was also investigated to select parameters for plant operators to monitor the onset of air entrainment into shutdown cooling system (SCS). Effects of the bellmouth entry on the critical submergence were also investigated. (author)

  18. General features of direct-cycle, supercritical-pressure, light-water-cooled reactors

    International Nuclear Information System (INIS)

    The concept of direct-cycle, supercritical-pressure, light-water-cooled reactors is developed. Breeding is possible in the tight lattice core. The power output can be maximized in the fast converter reactor. The gross thermal efficiency of the high temperature reactor adopting Inconel as fuel cladding is expected to be 44.8%. The plant system is similar to the supercritical-fossil-fired power plant which adopts once-through type coolant circulation system. The volume and height of the containment are approximately half of the BWR. The basic safety principles follows those of LWRs. The reactor will solve the economic problems of LWR and LMFBR

  19. 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.

  20. Performance comparison of metallic, actinide burning fuel in lead-bismuth and sodium cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, K.D.; Herring, J.S.; Macdonald, P.E. [Idaho National Engineering and Environment Lab., Advanced Nuclear Energy, Idaho (United States)

    2001-07-01

    Various methods have been proposed to ''incinerate'' or ''transmute'' the current inventory of transuranic waste (TRU) that exits in spent light-water-reactor (LWR) fuel, and weapons plutonium. These methods include both critical (e.g., fast reactors) and non-critical (e.g., accelerator transmutation) systems. The work discussed here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead and lead-alloy cooled fast reactors for producing low-cost electricity as well as for actinide burning. The neutronics of non fertile fuel loaded with 20 or 30-wt% light water reactor (LWR) plutonium plus minor actinides for use in a lead-bismuth cooled fast reactor are discussed in this paper, with an emphasis on the fuel cycle life and isotopic content. Calculations show that the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from -1.02 to -1.16 g/MWd, compared to a typical LWR actinide generation rate of 0.303 g/MWd. However, when using the same parameters, the sodium-cooled case went subcritical after 0.2 to 0.8 effective full power years, and the lead-bismuth cooled case ranged from 1.5 to 4.5 effective full power years. (author)

  1. Performance Comparison of Metallic, Actinide Burning Fuel in Lead-Bismuth and Sodium Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, Kevan Dean; Herring, James Stephen; Mac Donald, Philip Elsworth

    2001-04-01

    Various methods have been proposed to “incinerate” or “transmutate” the current inventory of trans-uranic waste (TRU) that exits in spent light-water-reactor (LWR) fuel, and weapons plutonium. These methods include both critical (e.g., fast reactors) and non-critical (e.g., accelerator transmutation) systems. The work discussed here is part of a larger effort at the Idaho National Engineering and Environmental Laboratory (INEEL) and at the Massachusetts Institute of Technology (MIT) to investigate the suitability of lead and lead-alloy cooled fast reactors for producing low-cost electricity as well as for actinide burning. The neutronics of non-fertile fuel loaded with 20 or 30-wt% light water reactor (LWR) plutonium plus minor actinides for use in a lead-bismuth cooled fast reactor are discussed in this paper, with an emphasis on the fuel cycle life and isotopic content. Calculations show that the average actinide burn rate is similar for both the sodium and lead-bismuth cooled cases ranging from -1.02 to -1.16 g/MWd, compared to a typical LWR actinide generation rate of 0.303 g/MWd. However, when using the same parameters, the sodium-cooled case went subcritical after 0.2 to 0.8 effective full power years, and the lead-bismuth cooled case ranged from 1.5 to 4.5 effective full power years.

  2. Coated particle fuel for high temperature gas cooled reactors

    International Nuclear Information System (INIS)

    for process heat/hydrogen generation applications with 950 .deg. C outlet temperatures. There is a clear set of standards for modern high quality fuel in terms of low levels of heavy metal contamination, manufacture-induced particle defects during fuel body and fuel element making, irradiation/accident induced particle failures and limits on fission product release from intact particles. While gas-cooled reactor design is still open-ended with blocks for the prismatic and spherical fuel elements for the pebble-bed design, there is near worldwide agreement on high quality fuel: a 500 μm diameter UO2 kernel of 10% enrichment is surrounded by a 100 μm thick sacrificial buffer layer to be followed by a dense inner pyrocarbon layer, a high quality silicon carbide layer of 35 μm thickness and theoretical density and another outer pyrocarbon layer. Good performance has been demonstrated both under operational and under accident conditions, i.e. to 10% FIMA and maximum 1600 .deg. C afterwards. And it is the wide-ranging demonstration experience that makes this particle superior. Recommendations are made for further work: 1. Generation of data for presently manufactured materials, e.g. SiC strength and strength distribution, PyC creep and shrinkage and many more material data sets. 2. Renewed start of irradiation and accident testing of modern coated particle fuel. 3. Analysis of existing and newly created data with a view to demonstrate satisfactory performance at burnups beyond 10% FIMA and complete fission product retention even in accidents that go beyond 1600 .deg. C for a short period of time. This work should proceed at both national and international level

  3. Thermal-hydraulic simulation and analysis of Research Reactor Cooling Systems

    International Nuclear Information System (INIS)

    The objective of the present study is to formulate a model to simulate the thermal hydraulic behavior of integrated cooling system in a typical material testing reactor (MTR) under loss of ultimate heat sink, the model involves three interactively coupled sub-models for reactor core, heat exchanger and cooling tower. The developed model predicts the temperature profiles in addition it predicts inlet and outlet temperatures of the hot and cold stream as well as the heat exchangers and cooling tower. The model is validated against PARET code for steady-state operation and also verified by the reactor operational records, and then the model is used to simulate the thermal-hydraulic behavior of the reactor under a loss of ultimate heat sink. The simulation is performed for two operational regimes named regime I of (11 MW) thermal power and three operated cooling tower cells and regime II of (22 MW) thermal power and six operated cooling tower cells. In regime I, the simulation is performed for 1, 2 and 3 cooling tower failed cells while in regime II, it is performed for 1, 2, 3, 4, 5 and 6 cooling tower failed cells. The safety action is conducted by the reactor protection system (RPS) named power reduction safety action, it is triggered to decrease the reactor power by amount of 20% of the present power when the water inlet temperature to the core reaches 43 degree C and a scram (emergency shutdown) is triggered in case of the inlet temperature reaches 44 degree C. The model results are analyzed and discussed. The temperature profiles of fuel, clad and coolant are predicted during transient where its maximum values are far from thermal hydraulic limits.

  4. 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.

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

    International Nuclear Information System (INIS)

    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 and D) that will be critical to the success of the NGNP, primarily in the areas of: (1) High temperature gas reactor fuels behavior; (2) High temperature materials qualification; (3) Design methods development and validation; (4) Hydrogen production technologies; and (5) Energy conversion. This paper presents current R and D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs

  6. Thermal hydraulics of the very high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    The Idaho National Laboratory (INL), under the auspices of the U.S. Department of Energy, 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 and 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 and D work that addresses fundamental thermal hydraulics issues that are relevant to a variety of possible NGNP designs. (author)

  7. 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)

  8. Air cooling of disk of a solid integrally cast turbine rotor for an automotive gas turbine

    Science.gov (United States)

    Gladden, H. J.

    1977-01-01

    A thermal analysis is made of surface cooling of a solid, integrally cast turbine rotor disk for an automotive gas turbine engine. Air purge and impingement cooling schemes are considered and compared with an uncooled reference case. Substantial reductions in blade temperature are predicted with each of the cooling schemes studied. It is shown that air cooling can result in a substantial gain in the stress-rupture life of the blade. Alternatively, increases in the turbine inlet temperature are possible.

  9. Shutdown cooling helium circulator design considerations for MHTGR [Modular High Temperature Gas-Cooled Reactor] power plant

    International Nuclear Information System (INIS)

    The Modular High Temperature Gas-Cooled Reactor (MHTGR) plant embodies a shutdown cooling system to expedite plant cooldown for refueling, maintenance, and repair in the event that the main cooling loop is unavailable. This is a non safety related system. A key component in this system, is a helium circulator. Oriented vertically, the rotating assembly in this machine is supported on active magnetic bearings, and the radial flow compressor is driven by a submerged induction electric motor rated at 160 kW(e). This paper gives details of the circulator design considerations and includes topics related to the machine operation and maintenance, and the technology base. 12 refs., 11 figs., 3 tabs

  10. CFD modeling and thermal-hydraulic analysis for the passive decay heat removal of a sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Research highlights: → The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maamora (CENM), Morocco. → The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW). → The most important conclusion is that all obtained values of DNBR, fuel center and surface temperature, cladding surface temperature and coolant temperature across the hottest channel are largely far to compromise safety of the reactor. - Abstract: In this study, a pool-typed design similar to sodium-cooled fast reactor (SFR) of the fourth generation reactors has been modeled using CFD simulations to investigate the characteristics of a passive mechanism of Shutdown Heat Removal System (SHRS). The main aim is to refine the reactor pool design in terms of temperature safety margin of the sodium pool. Thus, an appropriate protection mechanism is maintained in order to ensure the safety and integrity of the reactor system during a shutdown mode without using any active heat removal system. The impacts on the pool temperature are evaluated based on the following considerations: (1) the aspect ratio of pool diameter to depth, (2) the values of thermal emissivity of the surface materials of reactor and guard vessels, and (3) innerpool liner and core periphery structures. The computational results show that an optimal pool design in geometry can reduce the maximum pool temperature down to ∼551 oC which is substantially lower than ∼627 oC as calculated for the reference case. It is also concluded that the passive Reactor Air Cooling System (RACS) is effective in removing decay heat after shutdown. Furthermore, thermal radiation from the surface of the reactor vessel is found to be important; and thus, the selection of the vessel surface materials with a high emissivity would be a

  11. CFD modeling and thermal-hydraulic analysis for the passive decay heat removal of a sodium-cooled fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hung, T.C., E-mail: tchung@ntut.edu.t [Department of Mechanical Engineering, National Taipei University of Technology, 1, Sec. 3, Chung-hsiao E. Rd., Taipei 10608, Taiwan (China); Dhir, V.K. [Department of Mechanical and Aerospace Engineering, UCLA, CA (United States); Chang, J.C. [Graduate Institute of Mechanical and Electrical Engineering, National Taipei University of Technology, Taiwan (China); Wang, S.K. [Department of Mechanical and Automation Engineering, I-Shou University, Taiwan (China)

    2011-01-15

    Research highlights: The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maamora (CENM), Morocco. The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW) The most important conclusion is that all obtained values of DNBR, fuel center and surface temperature, cladding surface temperature and coolant temperature across the hottest channel are largely far to compromise safety of the reactor. - Abstract: In this study, a pool-typed design similar to sodium-cooled fast reactor (SFR) of the fourth generation reactors has been modeled using CFD simulations to investigate the characteristics of a passive mechanism of Shutdown Heat Removal System (SHRS). The main aim is to refine the reactor pool design in terms of temperature safety margin of the sodium pool. Thus, an appropriate protection mechanism is maintained in order to ensure the safety and integrity of the reactor system during a shutdown mode without using any active heat removal system. The impacts on the pool temperature are evaluated based on the following considerations: (1) the aspect ratio of pool diameter to depth, (2) the values of thermal emissivity of the surface materials of reactor and guard vessels, and (3) innerpool liner and core periphery structures. The computational results show that an optimal pool design in geometry can reduce the maximum pool temperature down to {approx}551 {sup o}C which is substantially lower than {approx}627 {sup o}C as calculated for the reference case. It is also concluded that the passive Reactor Air Cooling System (RACS) is effective in removing decay heat after shutdown. Furthermore, thermal radiation from the surface of the reactor vessel is found to be important; and thus, the selection of the vessel surface materials with a high emissivity

  12. 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.

  13. 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.

  14. 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.

  15. An evolutionary approach to advanced water cooled reactors

    International Nuclear Information System (INIS)

    Based on the result of the Feasibility Study undertaken since 1991, Indonesia may enter in the new nuclear era by introduction of several Nuclear Power Plants in our energy supply system. Requirements for the future NPP's are developed in two step approach. First step is for the immediate future that is the next 50 years where the system will be dominated by A-LWR's/A-PHWR's and the second step is for the time period beyond 50 years in which new reactor systems may start to dominate. The integral reactor concept provides a revolutionary improvements in terms of conceptual and safety. However, it creates a new set of complex machinery and operational problems of its own. The paper concerns with a brief description of nuclear technology status in Indonesia and a qualitative assessment of integral reactor concept. (author)

  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. Status and Future Challenges of CFD for Liquid Metal Cooled Reactors

    International Nuclear Information System (INIS)

    Liquid metal cooled reactors are envisaged to play an important role in the future of nuclear energy production because of their possibility to use natural resources efficiently and to reduce the volume and lifetime of nuclear waste. Typically, sodium and lead(-alloys) are envisaged as coolants for such reactors. Obviously, in the development of these reactors, thermal-hydraulics is recognized as a key (safety) challenge. A relatively new technique to deal with thermal-hydraulics issues is Computational Fluid Dynamics (CFD). This technique is used increasingly nowadays for design and safety evaluation purposes. This paper will discuss the development status of CFD application to liquid metal cooled reactors. In addition, the main challenges for future developments will be indicated. Firstly, the technological challenges will be discussed which ask for CFD application. Afterwards, the needs for CFD development and/or validation will be discussed. The discussion will also include the need for accompanying experiments. (author)

  18. 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.

  19. Verification calculations as per CFD FLOWVISION code for sodium-cooled reactor plants

    International Nuclear Information System (INIS)

    The paper studies the experience in application of CFD FlowVision software for analytical validation of sodium-cooled fast reactor structure components and the results of performed verification, namely: – development and implementation of new model of turbulent heat transfer in liquid sodium (LMS) in FlowVision software and model verification based on thermohydraulic characteristics studied by experiment at TEFLU test facility; – simulation of flowing and mixing of coolant with different temperatures in the upper mixing chamber of fast neutron reactor through the example of BN-600 (comparison with the results obtained at the operating reactor). Based on the analysis of the results obtained, the efficiency of CFD codes application for the considered problems is shown, and the proposals for CFD codes verification development as applied to the advanced sodium-cooled fast reactor designs are stated. (author)

  20. Thermal hydraulics of sodium-cooled fast reactors - key issues and highlights

    International Nuclear Information System (INIS)

    In this paper key issues and highlighted topics in thermal hydraulics are discussed in connection to the current Japan's sodium-cooled fast reactor development efforts. In particular, design study and related researches of the Japan Sodium-cooled Fast Reactor (JSFR) are focused. Several innovative technologies, e.g., compact reactor vessel, two-loop system, fully natural circulation decay heat removal, and recriticality free core, have been investigated in order to reduce construction cost and to achieve higher level of reactor safety. Preliminary evaluations of innovative technologies to be applied to JSFR are on-going. Here, progress of design study is introduced. Then, research and development activities on the thermal hydraulics related to the innovative technologies are briefly reviewed. (author)

  1. Progress in development and design aspects of advanced water cooled reactors

    International Nuclear Information System (INIS)

    The objective of the Technical Committee Meeting (TCM) was to provide an international forum for technical specialists to review and discuss technology developments and design work for advanced water cooled reactors, safety approaches and features of current water cooled reactors and to identify, understand and describe advanced features for safety and operational improvements. The TCM was attended by 92 participants representing 18 countries and two international organizations and included 40 presentations by authors of 14 countries and one international organization. A separate abstract was prepared for each of these presentations. Refs, figs, tabs

  2. 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.

  3. Preliminary Investigation of an Optimally Scramming Control Rod for Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    A passively safe control rod for gas-cooled reactors is proposed. This Optimally Scramming Control Rod (OSCR) is lifted out of the core region by the core coolant and descends back into the core when the coolant flow is not sufficient for core cooling purposes or in the event of depressurization. It is shown that for the current design of the OSCR, the reactor can be operated under normal lower power conditions down to about 80% of total power. It is also shown that cold shutdown can be achieved with rods of sufficiently low mass to allow naturally passive operation of the concept. (authors)

  4. Methods and technologies for cost reduction in the design of water cooled reactor power plants

    International Nuclear Information System (INIS)

    The Specialists Meeting was organized in the framework of the IAEA International Working Group on Advanced Technologies for Water-Cooled Reactors. Its purpose was to provide an international forum for review and discussion on recent results in research and development on different methods and technologies of current and advanced water-cooled reactor power plants, which can lead to reduced investment and operation, maintenance and fuel-cycle costs of the plants. 27 specialists representing 10 countries and the IAEA took part in the meeting. 10 papers were presented. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  5. 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.

  6. Improving economics and safety of water cooled reactors. Proven means and new approaches

    International Nuclear Information System (INIS)

    Nuclear power plants (NPPs) with water cooled reactors [either light water reactors (LWRs) or heavy water reactors (HWRs)] constitute the large majority of the currently operating plants. Water cooled reactors can make a significant contribution to meeting future energy needs, to reducing greenhouse gas emissions, and to energy security if they can compete economically with fossil alternatives, while continuing to achieve a very high level of safety. It is generally agreed that the largest commercial barrier to the addition of new nuclear power capacity is the high capital cost of nuclear plants relative to other electricity generating alternatives. If nuclear plants are to form part of the future generating mix in competitive electricity markets, capital cost reduction through simplified designs must be an important focus. Reductions in operating, maintenance and fuel costs should also be pursued. The Department of Nuclear Energy of the IAEA is examining the competitiveness of nuclear power and the means for improving its economics. The objective of this TECDOC is to emphasize the need, and to identify approaches, for new nuclear plants with water cooled reactors to achieve competitiveness while maintaining high levels of safety. The cost reduction methods discussed herein can be implemented into plant designs that are currently under development as well as into designs that may be developed in the longer term. Many of the approaches discussed also generally apply to other reactor types (e.g. gas cooled and liquid metal cooled reactors). To achieve the largest possible cost reductions, proven means for reducing costs must be fully implemented, and new approaches described in this document should be developed and implemented. These new approaches include development of advanced technologies, increased use of risk-informed methods for evaluating the safety benefit of design features, and international consensus regarding commonly acceptable safety requirements that

  7. 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.

  8. Improvement of the Decay Heat Removal Characteristics of the Generation IV Gas-cooled Fast Reactor

    OpenAIRE

    Epiney, Aaron Simon

    2010-01-01

    Gas cooling in nuclear power plants (NPPs) has a long history, the corresponding reactor types developed in France, the UK and the US having been thermal neutron-spectrum systems using graphite as the moderator. The majority of NPPs worldwide, however, are currently light water reactors, using ordinary water as both coolant and moderator. These NPPs – of the so-called second generation – will soon need replacement, and a third generation is now being ...

  9. Sustainability and Efficiency Improvements of Gas-Cooled High Temperature Reactors

    OpenAIRE

    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 future high-tech process heat applications with minimized R&D. The development of HTR started in the 1950s as a graphite moderated and helium cooled reactor. This concept featured important inherent...

  10. The key device--elevator in 10 MW high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    The basic structure, working principle and behavior of the control system of the elevator in 10 MW high temperature gas-cooled reactor (HTR-10) are researched. The five-phase hybrid stepping motor and the closed-loop control are adopted in the construction design of the elevator. About 20000 fuel elements and graphite balls were transported into the reactor core by the elevator to achieve the critical loading for HTR-10

  11. Sodium experiment on fully natural circulation systems for decay heat removal in Japan sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Fully natural circulation system is adopted in a decay heat removal system (DHRS) of Japan Sodium Cooled Fast Reactor (JSFR). The DHRS of JSFR consists of one unit of DRACS (direct reactor auxiliary cooling system), which has a dipped heat exchanger in the reactor vessel and two units of PRACS, which has a heat exchanger in a primary-side inlet plenum of IHX in each loop. Sodium experiments were carried out for heat transfer characteristics of a sodium-sodium heat exchanger of PRACS and start-up transient of the DHRS loop with parameters of pressure loss coefficients in the loops. The transient experiments for the start-up of DHRS loop showed that quick increase of natural draft in the air duct followed by smooth increase of sodium flow rate in the DHRS loop. Influences of the pressure loss coefficients in the primary loop and the DHRS loop were limited on the core temperature and also heat removal of PRACS, respectively due to recovery of natural circulation head via the increase of temperature difference in each loop. (author)

  12. 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.

  13. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-03-01

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

  15. 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)

  16. 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.

  17. Design guidelines for the forced-air cooling process of strawberries

    Energy Technology Data Exchange (ETDEWEB)

    Ferrua, M.J.; Singh, R.P. [Department of Biological and Agricultural Engineering, University of California, Davis, CA 95616 (United States)

    2009-12-15

    The complex structure of the package systems currently used by the strawberry industry has prevented an efficient design of the forced-air cooling process. In this study, we investigated the mechanisms through which different design parameters affect the rate and uniformity of cooling, using a previously validated computational fluid dynamics model. The results indicated that the vent area has a significant effect on the cooling rate, but not on its uniformity. A design that reduces bypassing will not necessarily increase the cooling rate, because there is less bypass air to cool down the air exiting from each clamshell, so that the air entering the next clamshell becomes warmer. Periodic airflow reversal improves the rate and homogeneity of the cooling process. (author)

  18. 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... RULEâ) Pt. 305, App. H Appendix H to Part 305—Cooling Performance and Cost for Central Air...

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

    International Nuclear Information System (INIS)

    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

  20. 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.

  1. 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

  2. Integration of High-Temperature Gas-Cooled Reactors into Industrial Process Applications

    Energy Technology Data Exchange (ETDEWEB)

    Lee Nelson

    2011-09-01

    This report is a summary of analyses performed by the NGNP project to determine whether it is technically and economically feasible to integrate high temperature gas cooled reactor (HTGR) technology into industrial processes. To avoid an overly optimistic environmental and economic baseline for comparing nuclear integrated and conventional processes, a conservative approach was used for the assumptions and calculations.

  3. The choice between cooled tubular reactor models: analysis of the hot spot

    NARCIS (Netherlands)

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

    1990-01-01

    The applicability of the one-dimensional pseudo-homogeneous model of the cooled tubular reactor is studied. Using the two-dimensional model as the more accurate one we compared both models by studying the influence of the design and operating variables on the conditions in the hot spot of the reacto

  4. 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)

  5. Neutron spectrum calculation and safety analysis for supercritical water-cooled reactor

    International Nuclear Information System (INIS)

    The supercritical water reactor is one of the six reactors recommended by Generation Ⅳ International Forum. Compared with existing light water reactors, the supercritical water reactor has advantages of high thermal efficiency, simplified system structure and low cost. The physical model of the supercritical water reactor is established with MCNP program in this paper, which solves the problem of intricate geometry of fuel assembly. The change of coolant density along the axis is considered and the neutron spectrum distribution of different regions of the core is calculated. The safety in loss of coolant accident for the supercritical water reactor and the effect of missing coolant in different regions on the reactivity and effective multiplication factor analyzed. The results show the supercritical water reactor core has high security. The countermeasures of loss of coolant accident is studied and the effectiveness of boron water cooling is validated. The research not only provide important reference for the construction and security analysis of the supercritical water reactor, but also has great significance for the application and development of the supercritical water reactor. (authors)

  6. 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.

  7. Status of liquid metal cooled fast breeder reactors

    International Nuclear Information System (INIS)

    This document represents a compilation of the information on the status of fast breeder reactor development. It is intended to provide complete and authoritative information for academic, energy, industrial and planning organizations in the IAEA Member States. The Report also provides extended reference and bibliography lists. A summarized overview of the national programmes of LMFBR development is given in Chapter II. Chapter III on LMFBR experience provides a brief description and purpose of all fast reactors - experimental, demonstration and commercial size - that have been or are planned for construction and operation. Fast reactor physics is dealt with in Chapter IV. Besides the basic facts and definitions of neutronics and the compilation and measurement of nuclear data, a broad range of the calculation methods, codes, and the state of the art is described. In Chapter V, fuels and materials are described. The emphasis is on the design and development experience gained with mixed oxide fuel pins and subassemblies. Structural materials, blanket elements and absorber materials are also discussed. Chaper VI presents a broad overview of the technical and engineering aspects of LMFBR power plants. LMFBR core design is described in detail, followed by the components of the main heat transport system, the refuelling equipment, and auxiliary systems. Chapter VII on safety is a compilation of the current safety design concepts of LMFBRs and new trends in safety criteria and safety goals. The chapter concludes with risk analyses of LMFBR technology. In Chapter VIII, the systems approach has been emphasized in the consideration of the whole LMFBR fuel cycle. Special emphasis is placed on safeguards aspects and the environmental impact of the LMFBR fuel cycle. Chapter IX describes deployment considerations of LMFBRs. Special emphasis is placed on economic aspects of the LMFBR power plant and its related fuel cycle. Finally, Chapter X provides an overall summary and a

  8. Failed fuel identification techniques for liquid-metal cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lambert, J.D.B.; Gross, K.C.; Mikaili, R. [Argonne National Lab., IL (United States); Frank, S.M.; Cutforth, D.C.; Angelo, P.L. [Argonne National Lab., Idaho Falls, ID (United States)

    1995-06-01

    The Experimental Breeder Reactor II (EBR-II), located in Idaho and operated for the US Department of Energy by Argonne National Laboratory, has been used as an irradiation testbed for LMR fuels and components for thirty years. During this time many endurance tests have been carried out with experimental LMR metal, oxide, carbide and nitride fuel elements, in which cladding failures were intentionally allowed to occur. This paper describes methods that have been developed for the detection, identification and verification of fuel failures.

  9. SSTAR: The U.S. Lead-Cooled Fast Reactor (LFR)

    Energy Technology Data Exchange (ETDEWEB)

    Smith, C F; Halsey, W G; Brown, N W; Sienicki, J J; Moisseytsev, A; Wade, D C

    2007-09-25

    It is widely recognized that the developing world is the next area for major energy demand growth, including demand for new and advanced nuclear energy systems. With limited existing industrial and grid infrastructures, there will be an important need for future nuclear energy systems that can provide small or moderate increments of electric power (10-700 MWe) on small or immature grids in developing nations. Most recently, the Global Nuclear Energy Partnership (GNEP) has identified, as one of its key objectives, the development and demonstration of concepts for small and medium sized reactors (SMRs) that can be globally deployed while assuring a high level of proliferation resistance. Lead-cooled systems offer several key advantages in meeting these goals. The small lead-cooled fast reactor concept known as the Small Secure Transportable Autonomous Reactor (SSTAR) reactor has been under ongoing development under the U.S. Generation IV Nuclear Energy Systems Initiative. It a system designed to provide energy security to developing nations while incorporating features to achieve nonproliferation aims, anticipating GNEP objectives. This paper presents the motivation for development of internationally deployable nuclear energy systems as well as a summary of one such system, SSTAR, which is the U.S. Generation IV Lead-cooled Fast Reactor system.

  10. Experimental feasibility study of radial injection cooling of three-pad radial air foil bearings

    Science.gov (United States)

    Shrestha, Suman K.

    Air foil bearings use ambient air as a lubricant allowing environment-friendly operation. When they are designed, installed, and operated properly, air foil bearings are very cost effective and reliable solution to oil-free turbomachinery. Because air is used as a lubricant, there are no mechanical contacts between the rotor and bearings and when the rotor is lifted off the bearing, near frictionless quiet operation is possible. However, due to the high speed operation, thermal management is one of the very important design factors to consider. Most widely accepted practice of the cooling method is axial cooling, which uses cooling air passing through heat exchange channels formed underneath the bearing pad. Advantage is no hardware modification to implement the axial cooling because elastic foundation structure of foil bearing serves as a heat exchange channels. Disadvantage is axial temperature gradient on the journal shaft and bearing. This work presents the experimental feasibility study of alternative cooling method using radial injection of cooling air directly on the rotor shaft. The injection speeds, number of nozzles, location of nozzles, total air flow rate are important factors determining the effectiveness of the radial injection cooling method. Effectiveness of the radial injection cooling was compared with traditional axial cooling method. A previously constructed test rig was modified to accommodate a new motor with higher torque and radial injection cooling. The radial injection cooling utilizes the direct air injection to the inlet region of air film from three locations at 120° from one another with each location having three axially separated holes. In axial cooling, a certain axial pressure gradient is applied across the bearing to induce axial cooling air through bump foil channels. For the comparison of the two methods, the same amount of cooling air flow rate was used for both axial cooling and radial injection. Cooling air flow rate was

  11. Mathematical modelling of performance of safety rod and its drive mechanism in sodium cooled fast reactor during scram action

    Energy Technology Data Exchange (ETDEWEB)

    Rajan Babu, V., E-mail: vrb@igcar.gov.in [Indira Gandhi Centre for Atomic Research, Department of Atomic Energy, Kalpakkam 603102, Tamil Nadu (India); Thanigaiyarasu, G. [Rajalakshmi Engineering College, Chennai 602105, Tamil Nadu (India); Chellapandi, P. [Indira Gandhi Centre for Atomic Research, Department of Atomic Energy, Kalpakkam 603102, Tamil Nadu (India)

    2014-10-15

    Highlights: • Mathematical modelling of dynamic behaviour of safety rod during scram action in fast reactor. • Effects of hydraulics, structural interaction and geometry on drop time of safety rod are understood. • Using simplified model, drop time can be assessed replacing detailed CFD analysis. • Sensitivities of the related parameters on drop time are understood. • Experimental validation qualifies the modelling and computer software developed. - Abstract: Performance of safety rod and its drive mechanism which are parts of shutdown systems in sodium cooled fast reactor (SFR) plays a major role in ensuring safe operation of the plant during all the design basis events. The safety rods are to be inserted into the core within a stipulated time during off-normal conditions of the reactor. Mathematical modelling of dynamic behaviour of a safety rod and its drive mechanism in a typical 500 MWe SFR during scram action is considered in the present study. A full-scale prototype system has undergone qualification tests in air, water and in sodium simulating the operating conditions in the reactor. In this paper, the salient features of the safety rod and its mechanism, details related to mathematical modelling and sensitivity of the parameters having influence on drop time are presented. The outcomes of the numerical analysis are compared with the experimental results. In this process, the mathematical model and the computer software developed are validated.

  12. Evolutionary water cooled reactors: Strategic issues, technologies and economic viability. Proceedings of a symposium

    International Nuclear Information System (INIS)

    Symposium on evolutionary water cooled reactors: Strategic issues, technologies and economic viability was intended for managers in utilities, reactor design organizations and hardware manufacturing companies and for government decision makers who need to understand technological advances and the potential of evolutionary water cooled reactors to contribute to near and medium term energy demands. The topics addressed include: strategic issues (global energy outlook, the role of nuclear power in sustainable energy strategies, power generation costs, financing of nuclear plant projects, socio-political factors and nuclear safety requirements); technological advances (instrumentation and control, means od improving prevention and mitigation of severe accidents, development of passive safety systems); keys to economic viability (simplification, standardization, advances in construction and project management, feedback of experience from utilities into new designs, and effective management of plant operation)

  13. Effects of Uncertainties in Lead Cross Section Data in Analysis of Lead Cooled and Reflected Reactors

    International Nuclear Information System (INIS)

    There are numerous uncertainties in the analyses of innovative reactor designs, arising from approximations used in the solution of the transport equation, and in nuclear data processing and cross section libraries generation. This paper describes: the problems encountered in the analysis of the lead cooled and reflected reactors; the new cross section data libraries developed to overcome these problems; and applications of these new data libraries to the Encapsulated Nuclear Heat Source (ENHS) core benchmark analysis. The ENHS is a new lead-bismuth or lead cooled novel reactor concept that is fuelled with metallic alloy of Pu, U and Zr, and is designed to operate for 20 effective full power years without refuelling and with very small burnup reactivity swing. The computational tool benchmarked include MOCUP, a coupled MCNP-4C and ORIGEN2.1 utility codes with MCNP data libraries based on the newest evaluations. (author)

  14. Status of advanced light water cooled reactor designs 1996

    International Nuclear Information System (INIS)

    The present report, which is significantly more comprehensive than the previously one, addresses the rationale and basic motivations that lead to a continuing development of nuclear technology, provides an overview of the world status of current LWRs, describes the present market situations, and identifies desired characteristics for future plants. The report also provides a detailed description of utility requirements that largely govern today's nuclear development efforts, the situation with regard to enhanced safety objectives, a country wise description of the development activities, and a technical description of the various reactor designs in a consistent format. The reactor designs are presented in two categories: (1) evolutionary concepts that are expected to be commercially available soon; and (2) innovative designs. The report addresses the main technical characteristics of each concept without assessing or evaluating them from a particular point of view (e.g. safety or economics). Additionally, the report identifies basic reference documents that can provide further information for detailed evaluations. The report closes with an outlook on future energy policy developments

  15. Evaluation of a Design Concept for the Combined Air-water Passive Cooling PAFS+

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Sung Won; Kwon, Taesoon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The APR+ system provides the Passive Auxiliary Feed-water System (PAFS) for the passive cooling capability. However, the current design requirement for working time for the PAFS is about 8 hours only. Thus, current working time of PAFS can not meet the required 72 hours cooling capability for the long term SBO situation. To meet the 72 hours cooling, the pool capacity should be almost 3∼4 times larger than that of current water cooling tank. In order to continue the PAFS operation for 72 hours, a new passive air-water combined cooling system is proposed. This paper provides the feasibility study on the combined passive air-water cooling system. Figure 1 and 2 show the conceptual difference of the PAFS and combined passive air-water cooling system, respectively. Simple performance evaluation of the passive air cooling heat exchanger has been conducted by the MARS calculation. For the postulated FLB scenario, 4800 heat exchanger tubes and 5 m/s air velocity are not sufficient to sustain the PCCT pool level for 72 hour cooling. Further works on the system design and performance enhancing plan are required to fulfill the 72 hours long term passive cooling.

  16. 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.

  17. The core design of ALFRED, a demonstrator for the European lead-cooled reactors

    International Nuclear Information System (INIS)

    Highlights: • The design for the lead fast reactor is conceived in a comprehensive approach. • Neutronic, thermal-hydraulic, and transient analyses show promising results. • The system is designed to withstand even design extension conditions accidents. • Activation products in lead, including polonium, are evaluated. - Abstract: The European Union has recently co-funded the LEADER (Lead-cooled European Advanced DEmonstration Reactor) project, in the frame of which the preliminary designs of an industrial size lead-cooled reactor (1500 MWth) and of its demonstrator reactor (300 MWth) were developed. The latter is called ALFRED (Advanced Lead-cooled Fast Reactor European Demonstrator) and its core, as designed and characterized in the project, is presented here. The core parameters have been fixed in a comprehensive approach taking into account the main technological constraints and goals of the system from the very beginning: the limiting temperature of the clad and of the fuel, the Pu enrichment, the achievement of a burn-up of 100 GWd/t, the respect of the integrity of the system even in design extension conditions (DEC). After the general core design has been fixed, it has been characterized from the neutronic point of view by two independent codes (MCNPX and ERANOS), whose results are compared. The power deposition and the reactivity coefficient calculations have been used respectively as input for the thermal-hydraulic analysis (TRACE, CFD and ANTEO codes) and for some preliminary transient calculations (RELAP, CATHARE and SIM-LFR codes). The results of the lead activation analysis are also presented (FISPACT code). Some issues of the core design are to be reviewed and improved, uncertainties are still to be evaluated, but the verifications performed so far confirm the promising safety features of the lead-cooled fast reactors

  18. Design study and R and D progress on Japan sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    This paper describes the progress of the design study and research and development (R and D) for the Japan Sodium-cooled Fast Reactor (JSFR) implemented in the 'Fast Reactor Cycle Technology Development (FaCT)' project. A sodium-cooled fast reactor with an electric power of 1,500 MWe is targeted for commercialization at around 2050, and a demonstration reactor assuming a power output from 500 to 750 MWe is planned to start operation at around 2025. R and D on innovative technologies to achieve economic competitiveness and enhance reliability and safety is carried out for the commercialization. A compact reactor vessel without a vessel wall cooling system is pursued in consideration of the wall thickness enough to resist the severest seismic condition. A two-loop cooling system with shortened high-chromium steel piping is a crucial feature, and studies on the hydraulics in the pipe elbow and the fabrication capability of the pipes are being carried out. A double-walled straight tube steam generator is investigated to enhance the reliability against sodium/water reaction, and developmental works are progressing, including the thermal-hydraulic design and trial manufacturing for components. Self-Actuated Shutdown System (SASS) is being developed with safety analysis of the applicability for JSFR and experimental demonstration in the experimental fast reactor JOYO. An advanced fuel handling system is pursued to enhance economic performance. In parallel with considering the necessity of studies on alternative technologies, discussion on whether the innovative technologies can be adopted for JSFR is in progress to be finalized in 2010. (author)

  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. COMMISION DEBUGGING OF AIR-COOLED ISLAND SYSTEMS FOR 600 MW AIR-COOLED UNITS%600 MW空冷机组空冷岛的调试

    Institute of Scientific and Technical Information of China (English)

    杨海生; 李路江; 吴瑞涛; 刘春报; 刘红霞

    2009-01-01

    The problems appeared in commission debugging of the air-cooled island systems for two 600 MW air-cooled units of Guodian Longshan Power Generation Co Ltd,such as undue fast vacuum drop in air-cooled island during start-up of said units after their shutdown in winter, the rapid rise of back-pressure in said air-cooled isrand due to full-load operation of all cooling air fans and air-leakage in the steam seal system, etc. , have been analysed, and corresponding preventive measures a-dopted for above-mentioned problems in the commission debugging process being given. Regarding to a part of problems, which hadn't been solved in debugging, some concrete recommendations have been put forward.%对国电河北龙山发电有限公司2×600 MW空冷机组空冷岛的调试中出现的问题进行了分析,如冬季停机后起动空冷岛真空下降过快,全部空冷风机全负荷运转,汽封系统漏空气空冷岛背压急剧上升等,给出了调试中对上述问题采取的相关防范措施,并对调试中部分未能解决的问题,提出了具体建议.

  1. 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

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

    International Nuclear Information System (INIS)

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

  3. 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)

  4. 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...

  5. Comparison of transient analysis of LBE-cooled fast reactor and ADS under loss of heat sink accident

    International Nuclear Information System (INIS)

    Highlights: • A natural circulation LBE-cooled research reactor is developed by Institute of Nuclear Energy Safety Technology. • The safety advantages for 10 MWth fast and ADS reactors under loss of heat sink were compared. • A neutronics and thermal-hydraulics coupled simulation code NTC was employed in this study. - Abstract: A natural circulation LBE-cooled research reactor developed by Institute of Nuclear Energy Safety Technology (INEST) is proposed and designed to be 10 MWth. The reactor has two operation modes, which are LBE-cooled critical fast reactor mode and LBE-cooled accelerator-driven sub-critical system (ADS) reactor mode. In this paper, protected loss of heat sink (PLOHS) and unprotected loss of heat sink (ULOHS) transient accidents were simulated for both the critical and ADS reactors above by NTC-2D code, aiming at the investigation of the safety advantages for the two reactor modes under this typical transient condition (loss of heat sink, LOHS). The results showed that during PLOHS for both the two modes, all the key parameters (core power, fuel, cladding and coolant temperatures in the hottest channel) decreased to very small values after the reactor scrammed, which meant the reactors under the two modes were both safe. For ULOHS, the fuel, cladding and coolant temperatures of the fast reactor increased smaller than those of the sub-critical one, which means the fast reactor had a better safety advantage under LOHS transient

  6. 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.

  7. 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.

    2016-04-01

    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.

  8. 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.

  9. 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

  10. 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...

  11. Optimum design of bipolar plates for separate air flow cooling system of PEM fuel cells stacks

    Science.gov (United States)

    Franco, Alessandro

    2015-12-01

    The paper discusses about thermal management of PEM fuel cells. The objective is to define criteria and guidelines for the design of the air flow cooling system of fuel cells stacks for different combination of power density, bipolar plates material, air flow rate, operating temperature It is shown that the optimization of the geometry of the channel permits interesting margins for maintaining the use of separate air flow cooling systems for high power density PEM fuel cells.

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

    International Nuclear Information System (INIS)

    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 concept, and it

  13. 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

  14. a Dosimetry Assessment for the Core Restraint of AN Advanced Gas Cooled Reactor

    Science.gov (United States)

    Thornton, D. A.; Allen, D. A.; Tyrrell, R. J.; Meese, T. C.; Huggon, A. P.; Whiley, G. S.; Mossop, J. R.

    2009-08-01

    This paper describes calculations of neutron damage rates within the core restraint structures of Advanced Gas Cooled Reactors (AGRs). Using advanced features of the Monte Carlo radiation transport code MCBEND, and neutron source data from core follow calculations performed with the reactor physics code PANTHER, a detailed model of the reactor cores of two of British Energy's AGR power plants has been developed for this purpose. Because there are no relevant neutron fluence measurements directly supporting this assessment, results of benchmark comparisons and successful validation of MCBEND for Magnox reactors have been used to estimate systematic and random uncertainties on the predictions. In particular, it has been necessary to address the known under-prediction of lower energy fast neutron responses associated with the penetration of large thicknesses of graphite.

  15. Designs for remote inspection of the ALMR Reactor Vessel Auxiliary Cooling System (RVACS)

    International Nuclear Information System (INIS)

    One of the most important safety systems in General Electric's (GI) Advanced Liquid Metal Reactor (ALMR) is the Reactor Vessel Auxiliary Cooling System (RVACS). Because of high temperature, radiation, and restricted space conditions, GI desired methods to remotely inspect the RVACS, emissive coatings, and reactor vessel welds during normal refueling operations. The DOE/NE Robotics for Advanced Reactors program formed a team to evaluate the ALMR design for remote inspection of the RVACS. Conceptual designs for robots to perform the required inspection tasks were developed by the team. Design criteria for these remote systems included robot deployment, power supply, navigation, environmental hardening of components, tether management, communication with an operator, sensing, and failure recovery. The operation of the remote inspection concepts were tested using 3-D simulation models of the ALMR. In addition, the team performed an extensive technology review of robot components that could survive the environmental conditions in the RVACS

  16. Dynamic response simulation for high temperature gas-cooled reactor with indirect closed Brayton cycle

    International Nuclear Information System (INIS)

    A transient simulation program is developed in order to study dynamic characteristics of high temperature gas-cooled reactor with indirect closed Brayton cycle. After the brief introduction to such a plant, detailed mathematical models for important installations are described in the paper. By inducing step positive reactivity into the reactor, it looks like that the powers of turbo machine installations have a different growth rate accompanied with small increase of reactor power. Furthermore, this paper shows the temperature changes of reactor and heat exchangers. For the heat exchangers of the whole secondary loop, the pressure changes behave quite differently for those three sections divided by turbine, low pressure compressor and high pressure compressor. For all these equipments, the simulation program gives reasonable results and is in accordance with dynamic characteristics of their own. (authors)

  17. Study of risk reduction by improving operation of reactor core isolation cooling system

    International Nuclear Information System (INIS)

    The Fukushima Daiichi nuclear power plant fell into a station blackout (SBO) due to the earthquake and tsunami in which most of the core cooling systems were disabled. In the units 2 and 3, water injection to the core was performed only by water injection system with turbine driven pumps. In particular, it is inferred from observed plant parameters that the reactor core isolation cooling system (RCIC) continued its operation much longer than it was originally expected (8 hours). Since the preparation of safety measures did not work, the reactor core damaged. With a view to reduce risk of station blackout events in a BWR by accident management, this study investigated the efficacy of operation procedures that takes advantage of RCIC which can be operated with only equipment inside reactor building and does not require an AC power source. The efficacy was assessed in this study by two steps. The first step is a thermal hydraulic analysis with the RETRAN3D code to estimate the potential extension of duration of core cooling by RCIC and the second step is the estimation of time required for recovery of off-site power from experiences at nuclear power stations under the 3.11 earthquake. This study showed that it is possible to implement more reliable measures for accident termination and to greatly reduce the risk of SBO by the installation of accident management measures with use of RCIC for extension of core cooling under SBO conditions. (author)

  18. Helium circulator design concepts for the modular high temperature gas-cooled reactor (MHTGR) plant

    International Nuclear Information System (INIS)

    Two helium circulators are featured in the Modular High-Temperature Gas-Cooled Reactor (MHTGR) power plant - (1) the main circulator, which facilitates the transfer of reactor thermal energy to the steam generator, and (2) a small shutdown cooling circulator that enables rapid cooling of the reactor system to be realized. The 3170 kW(e) main circulator has an axial flow compressor, the impeller being very similar to the unit in the Fort St. Vrain (FSV) plant. The 164 kW(e) shutdown cooling circulator, the design of which is controlled by depressurized conditions, has a radial flow compressor. Both machines are vertically oriented, have submerged electric motor drives, and embody rotors that are supported on active magnetic bearings. As outlined in this paper, both machines have been conservatively designed based on established practice. The circulators have features and characteristics that have evolved from actual plant operating experience. With a major goal of high reliability, emphasis has been placed on design simplicity, and both machines are readily accessible for inspection, repair, and replacement, if necessary. In this paper, conceptual design aspects of both machines are discussed, together with the significant technology bases. As appropriate for a plant that will see service well into the 21st century, new and emerging technologies have been factored into the design. Examples of this are the inclusion of active magnetic bearings, and an automated circulator condition monitoring system. (author). 18 refs, 20 figs, 13 tabs

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

    International Nuclear Information System (INIS)

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

  20. Mechanical Property and Its Comparison of Superalloys for High Temperature Gas Cooled Reactor

    International Nuclear Information System (INIS)

    Since structural materials for high temperature gas cooled reactor are used during long period in nuclear environment up to 1000 .deg. C, it is important to have good properties at elevated temperature such as mechanical properties (tensile, creep, fatigue, creep-fatigue), microstructural stability, interaction between metal and gas, friction and wear, hydrogen and tritium permeation, irradiation behavior, corrosion by impurity in He. Thus, in order to select excellent materials for the high temperature gas cooled reactor, it is necessary to understand the material properties and to gather the data for them. In this report, the items related to material properties which are needed for designing the high temperature gas cooled reactor were presented. Mechanical properties; tensile, creep, and fatigue etc. were investigated for Haynes 230, Hastelloy-X, In 617 and Alloy 800H, which can be used as the major structural components, such as intermediate heat exchanger (IHX), hot duct and piping and internals. Effect of He and irradiation on these structural materials was investigated. Also, mechanical properties; physical properties, tensile properties, creep and creep crack growth rate were compared for them, respectively. These results of this report can be used as important data to select superior materials for high temperature gas reactor

  1. Gas-cooled Fast Reactor (GFR) fuel and In-Core Fuel Management

    Energy Technology Data Exchange (ETDEWEB)

    Weaver, K.D.; Sterbentz, J. [Idaho National Engineering and Environmental Laboratory, P.O. Box 1625, Idaho Falls, Idaho 83415-3850 (United States); Meyer, M. [Argonne National Laboratory- West (United States); Lowden, R. [Oak Ridge National Laboratory (United States); Hoffman, E.; Wei, T.Y.C. [Argonne National Laboratory (United States)]. e-mail: weavkd@inel.gov

    2004-07-01

    The Gas-Cooled Fast Reactor (GCFR) has been chosen as one of six candidates for development as a Generation IV nuclear reactor based on: its ability to fully utilize fuel resources; minimize or reduce its own (and other systems) actinide inventory; produce high efficiency electricity; and the possibility to utilize high temperature process heat. Current design approaches include a high temperature (2 850 C) helium cooled reactor using a direct Brayton cycle, and a moderate temperature (550 C - 650 C) helium or supercritical carbon dioxide (S-CO{sub 2}) cooled reactor using direct or indirect Brayton cycles. These design choices have thermal efficiencies that approach 45% to 50%, and have turbomachinery sizes that are much more compact compared to steam plants. However, there are challenges associated with the GCFR, which are the focus of current research. This includes safety system design for decay heat removal, development of high temperature/high fluence fuels and materials, and development of fuel cycle strategies. The work presented here focuses on the fuel and preliminary in-core fuel management, where advanced ceramic-ceramic (cercer) dispersion fuels are the main focus, and average burnups to 266 M Wd/kg appear achievable for the reference Si C/(U,TRU)C block/plate fuel. Solid solution (pellet) fuel in composite ceramic clad (Si C/Si C) is also being considered, but remains as a backup due to cladding fabrication challenges, and high centerline temperatures in the fuel. (Author)

  2. Operation and Licensing of Mixed Cores in Water Cooled Reactors

    International Nuclear Information System (INIS)

    Nuclear fuel is a highly complex material that is subject to continuous development and is produced by a range of manufacturers. During operation of a nuclear power plant, the nuclear fuel is subject to extreme conditions of temperature, corroding environment and irradiation, and many different designs of fuel have been manufactured with differing fuel materials, cladding materials and assembly structure to ensure these conditions. The core of an operating power plant can contain hundreds of fuel assemblies, and where there is more than a single design of a fuel assembly in the core, whether through a change of fuel vendor, introduction of an improved design or for some other reason, the core is described as a mixed core. The task of ensuring that the different assembly types do not interact in a harmful manner, causing, for example, differing flow resistance resulting in under cooling, is an important part of ensuring nuclear safety. This report has compiled the latest information on the operational experience of mixed cores and the tools and techniques that are used to analyse the core operation and demonstrate that there are no safety related problems with its operation. This publication is a result of a technical meeting in 2011 and a series of consultants meetings

  3. Frequency and distribution of leakages in steam generators of gas-cooled reactors

    International Nuclear Information System (INIS)

    In gas cooled reactors with graphitic primary circuit structures - such as HTR, AGR or Magnox - the water ingress is an event of great safety concern. Water or steam entering the primary circuit react with the hot graphite and carbon-oxide and hydrogen are produced. As the most important initiating event a leak in a steam generator must be taken into account. From the safety point of view as well as for availability reasons it is necessary to construct reliable boilers. Thus the occurrence of a boiler leak should be a rare event. In the context of a probabilistic safety study for an HTR-Project much effort was invested to get information about the frequency and the size distribution of tube failures in steam generators of gas cooled reactors. The main data base was the boiler tube failure statistics of United Kingdom gas cooled reactors. The data were selected and applied to a modern HTR steam generator design. A review of the data showed that the failure frequency is not connected with the load level (pressures, temperatures) or with the geometric size of the heating surface of the boiler. Design, construction, fabrication, examination and operation conditions have the greatest influence an the failure frequency but they are practically not to be quantified. The typical leak develops from smallest size. By erosion effects of the entering water or steam it is enlarged to perhaps some mm2, then usually it is detected by moisture monitors. Sudden tube breaks were not reported in the investigated period. As a rule boiler leaks in gas cooled reactors are much more, rare then leaks in steam generators of light water reactors and fossil fired boilers. (author)

  4. Research towards ultrasonic systems to assist in-vessel manipulations in liquid metal cooled reactors

    International Nuclear Information System (INIS)

    We describe the state of the art of the research towards ultrasonic measurement methods for use in lead-bismuth cooled liquid metal reactors. Our current research activities are highly focused on specific tasks in the MYRRHA system, which is a fast spectrum research reactor cooled with the eutectic mixture of lead and bismuth (LBE) and is conceived as an accelerator driven system capable of operating in both sub-critical and critical mode. As liquid metal is opaque to light, normal visual feedback during fuel manipulations in the reactor vessel is not available and must therefore be replaced by a system that is not hindered by the opacity of the coolant. In this respect ultrasonic measurement techniques have been proposed and even developed in the past for operation in sodium cooled reactors. To our knowledge, no such systems have ever been deployed in lead based reactors and we are the first to have a research program in this direction as will be detailed in this paper. We give an overview of the acoustic properties of LBE and compare them with the properties of sodium and water to theoretically show the feasibility of ultrasonic systems operating in LBE. In the second part of the paper we discuss the results of the validation experiments in water and LBE. A typical scene is ultrasonically probed by a mechanical scanning system while the signals are processed to render a 3D visualization on a computer screen. It will become clear that mechanical scanning is capable of producing acceptable images but that it is a time consuming process that is not fit to solve the initial task to providing feedback during manipulations in the reactor vessel. That is why we propose to use several dedicated ultrasonic systems each adapted to a specific task and capable to provide real-time feedback of the ongoing manipulations, as is detailed in the third and final part of the paper. (authors)

  5. 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

  6. Steam generators and heat exchangers for gas-cooled reactors. Background and status in Switzerland

    International Nuclear Information System (INIS)

    The Swiss company Sulzer Brothers Ltd. built its first nuclear steam generator in 1961 for a CO2-cooled prototype reactor. Since then the Company has been involved in the planning, development and manufacture of steam generators for gas-cooled reactors, in particular for the French Magnox reactor program. In 1980 Sulzer delivered the 6-module steam generator for the German High Temperature Reactor Prototype THTR-300. The production of hardware was continuously accompanied and supported by extensive research and development activities. Experimental programs comprised thermohydraulic investigations related to the primary gas-side as well as to the secondary side and its two-phase-flow stability. In the area of high temperature materials thermal cycling tests were performed to analyse the fatigue of bimetallic welds under severe transients. Low cycle creep fatigue damage in tube bends and the wear and fretting characteristics of protective coatings on the helium side of hot tubes were investigated. Fabrication experiments for large helical heat exchangers served to extrapolate known manufacturing technology to commercial size HTGR units. In the frame of international GCR programs Switzerland participated in the Gas Breeder Reactor Association and the High Temperature Helium Turbine Project. For these projects Sulzer designed and developed steam generators, recuperators and primary coolers

  7. Statics and dynamics of a natural circulation cooled boiling water reactor. Doctoral thesis

    Energy Technology Data Exchange (ETDEWEB)

    Stekelenburg, A.J.C.

    1994-02-21

    Due to the strong interconnection of the various processes in the reactor vessel of a natural circulation cooled boiling water reactor (BWR), explaining the physics of both the statics and the dynamics of the Dodewaard reactor is not an easy task. In this thesis, the physics is studied through a combined experimental and theoretical investigation. The experiments are analyzed further with the use of the model, and the results of the model calculations provide ideas for new experiments. For an experimental study of the reactor behavior, measurement tools are required. Many relevant process variables are supplied by the power plant's data-logger, but a direct method for measuring the circulation flow rate is not available. Reactor behavior can be studied theoreticallly with the use of a complex computer code, based on a multi-node model. In this way, reliable results are obtained. In many cases, however, such a code is not easy to use, and the calculations require much computer time. Calculations based on a simple model have a lower reliability, but, as the model is clearer, provide more insight into the physics of the system. For this reason, a simple theoretical dynamical model for the main physical processes of the Dodewaard natural circulation cooled BWR is presented in the thesis.

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

    International Nuclear Information System (INIS)

    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

  9. 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.

  10. Modular high-temperature gas-cooled reactor core heatup accident simulations

    International Nuclear Information System (INIS)

    The design features of the modular high-temperature gas-cooled reactor (HTGR) have the potential to make it essentially invulnerable to damage from postulated core heatup accidents. Simulations of long-term loss-of-forced-convection (LOFC) accidents, both with and without depressurization of the primary coolant and with only passive cooling available to remove afterheat, have shown that maximum core temperatures stay below the point at which fuel failures and fission product releases are expected. Sensitivity studies also have been done to determine the effects of errors in the predictions due both to uncertainties in the modeling and to the assumptions about operational parameters. 4 refs., 5 figs

  11. Design of a large-scale, multi-purpose high temperature gas-cooled reactor system

    International Nuclear Information System (INIS)

    The trial design of a large-scale, multi-purpose high temperature gas-cooled reactor system is described on its three aspects: nuclear reactor, nuclear heat utilization, and safety. The system is a littoral iron and steel making plant employing a multi-purpose HTGR (heat output 3,000 MW) with helium gas temperature of 1,0000C; the capacity is about 6,300,000 tons of crude steel production per year. It consists of a direct reduction furnace for ore and an electric furnace, and also an electric power generating facility. (Mori, K.)

  12. High-temperature gas-cooled reactor (HTGR): long term program plan

    International Nuclear Information System (INIS)

    The FY 1980 effort was to investigate four technology options identified by program participants as potentially viable candidates for near-term demonstration: the Gas Turbine system (HTGR-GT), reflecting its perceived compatibility with the dry-cooling market, two systems addressing the process heat market, the Reforming (HTGR-R) and Steam Cycle (HTGR-SC) systems, and a more developmental reactor system, The Nuclear Heat Source Demonstration Reactor (NHSDR), which was to serve as a basis for both the HTGR-GT and HTGR-R systems as well as the further potential for developing advanced applications such as steam-coal gasification and water splitting

  13. 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.

  14. Fuel performance models for high-temperature gas-cooled reactor core design

    International Nuclear Information System (INIS)

    Mechanistic fuel performance models are used in high-temperature gas-cooled reactor core design and licensing to predict failure and fission product release. Fuel particles manufactured with defective or missing SiC, IPyC, or fuel dispersion in the buffer fail at a level of less than 5 x 10-4 fraction. These failed particles primarily release metallic fission products because the OPyC remains intact on 90% of the particles and retains gaseous isotopes. The predicted failure of particles using performance models appears to be conservative relative to operating reactor experience

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

    OpenAIRE

    Xingtuan Yang; Yanfei Sun; Huaiming Ju; Shengyao Jiang

    2014-01-01

    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 ca...

  16. The effects of aging on Boiling Water Reactor core isolation cooling system

    International Nuclear Information System (INIS)

    A study was performed to assess the effects of aging on the Reactor Core Isolation Cooling system in commercial Boiling Water Reactors. This study is part of the Nuclear Plant Aging Research program sponsored by the US Nuclear Regulatory Commission. The failure data, from national databases, as well as plant specific data were reviewed and analyzed to understand the effects of aging on the RCIC system. This analysis identified important components that should receive the highest priority in terms of aging management. The aging characterization provided information on the effects of aging on component failure frequency, failure modes, and failure causes

  17. RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors

    International Nuclear Information System (INIS)

    The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point

  18. RELAP5-3D code for supercritical-pressure, light-water-cooled reactors

    International Nuclear Information System (INIS)

    The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point. (author)

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

    International Nuclear Information System (INIS)

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

  20. 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.

  1. 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.

  2. Experimental evaluation of a direct air-cooled lithium bromide-water absorption prototype for solar air conditioning

    OpenAIRE

    González-Gil, A.; Izquierdo, M.; Marcos, J.D.; Palacios, E.

    2011-01-01

    Abstract A new direct air-cooled single-effect LiBr-H2O absorption prototype is described and proposed for use in solar cooling. As distinguishing aspects, it presents: an adiabatic absorber using flat-fan sheets; an air-cooling system that directly refrigerates both the condenser and the absorber and; the possibility of being operated also as a double-effect unit. A solar facility comprising a 48m2 field of flat-plate collectors was used to test the single-effect operation mode of...

  3. Experimental results of a direct air-cooled ammonia–lithium nitrate absorption refrigeration system

    International Nuclear Information System (INIS)

    Absorption thermal cooling systems driven by renewable energy are a viable option in order to reduce fossil fuel consumption and the associated emissions. This work shows the results of an air cooled absorption cooling prototype working with an ammonia–lithium nitrate mixture at high ambient temperatures. An absorption refrigeration system was designed and built. The prototype is a one stage ammonia–lithium nitrate air cooled chiller. The experimental system was instrumented to evaluate each component. This paper shows the operation conditions in the experimental unit as well as some of the heat loads encountered at different operating conditions. The system was operated successfully at ambient temperatures in the range of 25–35 °C. A series of test showed that even at ambient temperatures it can be operated at evaporator temperatures below 10 °C producing chilled water for air conditioning applications such as radiative cooling panels. The system proved to stabilize very quickly and no risk of crystallization was encountered so the first results are promising in order to continue with the development of a more advanced prototype. - Highlights: •Experimental results of a direct air-cooled ammonia–lithium nitrate system. •The prototype is a one stage ammonia–lithium nitrate air cooled chiller. •The absorption system was operated successfully at ambient temperatures. •Cooling loads of 4.5 kW were reached in the chilled water side

  4. Computer code for the analyses of reactivity initiated accident of heavy water moderated and cooled research reactor 'EUREKA-2D'

    International Nuclear Information System (INIS)

    Codes, such as EUREKA and EUREKA-2 have been developed to analyze the reactivity initiated accident for light water reactor. These codes could not be applied directly for the analyses of heavy water moderated and cooled research reactor which are different from light water reactor not only on operation condition but also on reactor kinetic constants. EUREKA-2D which is modified EUREKA-2 is a code for the analyses of reactivity initiated accident of heavy water research reactors. Following items are modified: 1) reactor kinetic constants. 2) thermodynamic properties of coolant. 3) heat transfer equations. The feature of EUREKA-2D and an example of analysis are described in this report. (author)

  5. 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.

  6. 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.

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

    International Nuclear Information System (INIS)

    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 >300C, even in winter; (2) a maximal 50C 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

  8. Modular high-temperature gas-cooled reactor simulation using parallel processors

    International Nuclear Information System (INIS)

    The MHPP (Modular HTGR Parallel Processor) code has been developed to simulate modular high-temperature gas-cooled reactor (MHTGR) transients and accidents. MHPP incorporates a very detailed model for predicting the dynamics of the reactor core, vessel, and cooling systems over a wide variety of scenarios ranging from expected transients to very-low-probability severe accidents. The simulation routines, which had originally been developed entirely as serial code, were readily adapted to parallel processing Fortran. The resulting parallelized simulation speed was enhanced significantly. Workstation interfaces are being developed to provide for user (''operator'') interaction. The benefits realized by adapting previous MHTGR codes to run on a parallel processor are discussed, along with results of typical accident analyses. 3 refs., 3 figs

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

    International Nuclear Information System (INIS)

    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.

  10. 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.

  11. 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.

  12. Optimization of operational water chemistry for supercritical-water cooled reactor

    International Nuclear Information System (INIS)

    The paper summaries the experimental results obtained within the project 'PRAMEK'. The project is focused on the study of the compatibility of the construction material of fossil-fueled supercritical water cooled power plants and water chemistry, that is currently used and optimization the dosing of the chemical species to the working circuit. The experience from the project enables to evaluate the water chemistry for Supercritical water cooled reactor (SCWR) and the transfer of the operational experience to the operation of the future nuclear power plant. The used materials are candidate for the SCWR and used in the industrial scale in the Ledvice power plant (fossil fuelled) with the supercritical parameters of the medium. It illustrates the future behaviour in the SCWR plant. The influence of the irradiation will be tested in future within the supercritical water loop in the reactor LVR-15. (author)

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

    International Nuclear Information System (INIS)

    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.

  14. 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.

  15. Economic performance of liquid-metal fast breeder reactor and gas-cooled fast reactor radial blankets

    International Nuclear Information System (INIS)

    The economic performance of the radial blanket of a liquid-metal fast breeder reactor (LMFBR) and a gas-cooled fast reactor (GCFR) has been studied based on the calculation of the net financial gain as well as the value of the levelized fuel cost. The necessary reactor physics calculations have been performed using the code CITATION, and the economic analysis has been carried out with the code ECOBLAN, which has been written for that purpose. The residence time of fuel in the blanket is the main variable of the economic analysis. Other parameters that affect the results and that have been considered are the value of plutonium, the price of heat, the effective cost of money, and the holdup time of the spent fuel before reprocessing. The results show that the radial blanket of both reactors is a producer of net positive income for a broad range of values of the parameters mentioned above. The position of the fuel in the blanket and the fuel management scheme applied affect the monetary gain. There is no significant difference between the economic performance of the blanket of an LMFBR and a GCFR

  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. Performance analysis of solar air cooled double effect LiBr/H2O absorption cooling system in subtropical city

    International Nuclear Information System (INIS)

    Highlights: • The meteorological data during the working period of air conditioning was measured. • The suitable working range of collector temperature of system was gotten. • The characteristic of hourly and monthly total efficiency of system were obtained. • The yearly performance of system was calculated. - Abstract: Due to the absence of cooling tower and independent on water, the air cooled solar double effect LiBr/H2O absorption cooling system is more convenient to be used in commercial building and household use. The performance with collector temperature is an important field for such system. The paper mainly deals with the performance with collector temperature for the solar air cooled double effect LiBr/H2O absorption cooling system in subtropical city. The parameters of system are: aperture area of collector array is 27 m2, tilted angle of collector with respect to the horizontal plane is 20 toward to south evaporator temperature is 5 °C and the cooling capacity is 20 kW. The simulation is based on the meteorological data of monthly typical day which was summarized from a year round measured data. A corresponding parametric model was developed. The hourly and average performance with the collector temperature for monthly typical day was obtained and discussed. It was found that the suitable working range of inlet temperature of collector is 110–130 °C to improve performance and lower the risk of crystallization. The difference of hourly total efficiency in 9:00–16:00 is less, and the monthly total efficiency from May to October is approximate. The yearly performance of system including total efficiency, cooling capacity per area of collector and solar fraction was given. Furthermore, the effect of effectiveness of heat exchanger and pressure drop on total efficiency and solar fraction was studied and compared. The paper can serve as a preliminary investigation of solar air cooled double effect LiBr/H2O absorption cooling system in

  18. A concept of self-completed fuel cycle based on lead-cooled nitride-fuel fast reactors

    International Nuclear Information System (INIS)

    A concept of nuclear energy total system was studied based on the nitride fuel cycle and inherent safety lead-cooled fast reactors. In the nitride fuel reprocessing, a new concept for pyrochemical method was proposed due to reducing fuel cycle cost. The present designed lead-cooled fast reactors have higher safety, economics and minor actinide transmutation efficiency than those of MOX-fuel fast reactors. The construction of 1500 MWt plant is feasible as a result for technology studies for aseismic, steam-generator and reactor configuration systems. (author)

  19. A global model for gas cooled reactors for the Generation-4: application to the Very High Temperature Reactor (VHTR)

    International Nuclear Information System (INIS)

    Gas cooled high temperature reactor (HTR) belongs to the new generation of nuclear power plants called Generation IV. The Generation IV gathers the entire future nuclear reactors concept with an effective deployment by 2050. The technological choices relating to the nature of the fuel, the moderator and the coolant as well as the annular geometry of the core lead to some physical characteristics. The most important of these characteristics is the very strong thermal feedback in both active zone and the reflectors. Consequently, HTR physics study requires taking into account the strong coupling between neutronic and thermal hydraulics. The work achieved in this Phd consists in modeling, programming and studying of the neutronic and thermal hydraulics coupling system for block type gas cooled HTR. The coupling system uses a separate resolution of the neutronic and thermal hydraulics problems. The neutronic scheme is a double level Transport (APOLLO2) /Diffusion (CRONOS2) scheme respectively on the scale of the fuel assembly and a reactor core scale. The thermal hydraulics model uses simplified Navier Stokes equations solved in homogeneous porous media in code CAST3M CFD code. A generic homogenization model is used to calculate the thermal hydraulics parameters of the porous media. A de-homogenization model ensures the link between the porous media temperatures of the temperature defined in the neutronic model. The coupling system is made by external procedures communicating between the thermal hydraulics and neutronic computer codes. This Phd thesis contributed to the Very High Temperature Reactor (VHTR) physics studies. In this field, we studied the VHTR core in normal operating mode. The studies concern the VHTR core equilibrium cycle with the control rods and using the neutronic and thermal hydraulics coupling system. These studies allowed the study of the equilibrium between the power, the temperature and Xenon. These studies open new perspective for core

  20. Radiation Heat Transfer Effect on Thermal Sizing of Air-Cooling Heat Exchanger of Emergency Cooldown Tank

    International Nuclear Information System (INIS)

    An attempt has begun to extend the life time of emergency cooldown tank (ECT) by Korea Atomic Energy Research Institute (KAERI) researchers. Moon et al. recently reported a basic concept upon how to keep the ECT in operation beyond 72 hours after an accident occurs without any active corrective actions for the postulated design basis accidents. When the SMART (System-integrated Modular Advanced Reac-Tor) received its Standard Design Approval (SDA) for the first time in the world, hybrid safety systems are applied. However, the passive safety systems of SMART are being enforced in response to the public concern for much safer reactors since the Fukushima accident occurred. The ECT is a major component of a passive residual heat removal system (PRHRS), which is one of the most important systems to enhance the safety of SMART. It is being developed in a SMART safety enhancement project to contain enough cooling water to remove a sensible heat and a decay heat from reactor core for 72 hours since an accident occurs. Moon et al. offered to install another heat exchanger above the ECT and to recirculate an evaporated steam into water, which enables the ECT to be in operation, theoretically, indefinitely. An investigation was made to determine how long and how many tubes were required to meet the purpose of the study. In their calculation, however, a radiation heat transfer effect was neglected. The present study is to consider the radiation heat transfer for the design of air-cooling heat exchanger. Radiation heat transfer is normally ignored in many situations, but this is not the case for the present study. Kim et al. conducted thermal sizing of scaled-down ECT heat exchanger, which will be used to validate experimentally the basic concept of the present study. Their calculation is also examined to see if a radiation heat transfer effect was taken into consideration. The thermal sizing of an air-cooling heat exchanger was conducted including radiation heat transfer

  1. Analysis of passive residual heat removal system of modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    The passive residual heat removal system plays an important role for the inherent safety of high temperature gas-cooled reactor (HTGR). The thermal hydraulic calculation method for the residual heat removal system of HTGR was introduced. The operating temperatures of the residual heat removal system at different residual heat powers and different environmental temperatures were calculated. The containment concrete temperature was numerically simulated. The results show that the highest concrete temperature is acceptable. (authors)

  2. Mechanical properties of structural materials for high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Structural materials for high temperature gas cooled reactor should have good properties such as mechanical properties (tensile, creep, fatigue, creep-fatigue), microstructural stability, interaction between metal and gas, friction and wear, hydrogen and tritium permeation, irradiation behavior, corrosion by impurity in He. Mechanical properties of major structural materials, such as pressure vessel, heat exchanger, control rod, were investigated. Effect of He and irradiation on these structural materials were investigated

  3. Design features facilitating the decommissioning of advanced gas-cooled reactors

    International Nuclear Information System (INIS)

    The design of the advanced gas-cooled reactors is discussed as is the proposed decommissioning plan for delayed decommissioning. The special features which assist in decommissioning are presented. As a result of the study a catalogue of design features which will facilitate decommissioning is given. In addition to the catalogue of design features, the radioactive inventory 10 years after shutdown and 100 years after shutdown has been calculated. From this a provisional operator dose from activities associated with decommissioning has been assessed

  4. Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment

    International Nuclear Information System (INIS)

    Topics covered during the 'Specialists' meeting on gas-cooled reactor fuel development and spent fuel treatment' were as follows: Selection of constructions and materials, fuel element development concepts; Fabrication of spherical coated fuel particles and fuel element on their base; investigation of fuel properties; Spent fuel treatment and storage; Head-end processing of HTGR fuel elements; investigation of HTGR fuel regeneration process; applicability of gas-fluorine technology of regeneration of spent HTGR fuel elements

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

    International Nuclear Information System (INIS)

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

  6. Fabrication of spherical fuel element for 10 MW high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Cold quasi-isostatic molding with a silicon rubber die was used for manufacturing the spherical fuel elements of 10 MW high temperature gas-cooled reactor. 44 batches of fuel elements, about 20540 of the fuel elements, were produced. The cold properties of the graphite matrix materials satisfies the design specifications. The mean free uranium fraction in spherical fuel element from 44 batches is 4.57 x 10-5, certified products is 99%

  7. Preliminary Evaluation of a Nuclear Scenario Involving Innovative Gas Cooled Reactors

    OpenAIRE

    Eugene Shwageraus; Vincenzo Romanello; Guglielmo Lomonaco; Emil Fridman; Giuseppe Forasassi; Nicola Cerullo; Barbara Vezzoni

    2009-01-01

    In order to guarantee a sustainable supply of future energy demand without compromising the environment, some actions for a substantial reduction of C O 2 emissions are nowadays deeply analysed. One of them is the improvement of the nuclear energy use. In this framework, innovative gas-cooled reactors (both thermal and fast) seem to be very attractive from the electricity production point of view and for the potential industrial use along the high temperature processes (e.g., H 2 production b...

  8. Study plan for conducting a section 316(a) demonstration: K-Reactor cooling tower, Savannah River Site

    International Nuclear Information System (INIS)

    The K Reactor at the Savannah River Site (SRS) began operation in 1954. The K-Reactor pumped secondary cooling water from the Savannah River and discharged directly to the Indian Grave Branch, a tributary of Pen Branch which flows to the Savannah River. During earlier operations, the temperature and discharge rates of cooling water from the K-reactor were up to approximately 70 degree C and 400 cfs, substantially altering the thermal and flow regimes of this stream. These discharges resulted in adverse impacts to the receiving stream and wetlands along the receiving stream. As a component of a Consent Order (84-4-W as amended) with the South Carolina Department of Health and Environmental Control, the Department of Energy (DOE) evaluated the alternatives for cooling thermal effluents from K Reactor and concluded that a natural draft recirculating cooling tower should be constructed. The cooling tower will mitigate thermal and flow factors that resulted in the previous impacts to the Indian Grave/Pen Branch ecosystem. The purpose of the proposed biological monitoring program is to provide information that will support a Section 316(a) Demonstration for Indian Grave Branch and Pen Branch when K-Reactor is operated with the recirculating cooling tower. The data will be used to determine that Indian Grave Branch and Pen Branch support Balanced Indigenous Communities when K-Reactor is operated with a recirculating cooling tower. 4 refs., 1 fig. 1 tab

  9. Challenges and Innovative Technologies on Fuel Handling Systems for Future Sodium Cooled Fast Reactors

    International Nuclear Information System (INIS)

    Fast Reactors have a unique capability as a sustainable energy source in terms of both utilisation of fissile material for energy production and minimisation of the nuclear waste, due to the hard neutron spectrum. As a result of a screening review of candidate technologies and in the frame of the international forum Generation IV, Sodium Fast Reactors (SFR) are amongst the selected systems to address the sustainability issues with a coherent set of innovative requirements. The guidelines for the definition of such innovative requirements are the Generation IV goals with significant improvements on economy, safety, environment, waste management and proliferation resistance as promising milestone towards a sustainable nuclear energy. CEA, AREVA and EDF have an extensive experience and significant expertise in Sodium cooled Fast Reactors over the past 40 years of R and D and feedback experiments. Some improvements are needed on the SFR to meet the GEN IV goals, and in particular the reduction of investment and operating costs: the Fuel Handling System (FHS) can be considered as an essential step in the reactor design. The reactor refuelling system provides the means of transporting, storing and handling reactor core subassemblies. The system consists of the facilities and equipments 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 account various items and in particular operating strategies such as core design and management and core configuration. Moreover, the FHS will have to cope with safety assessments: a permanent cooling strategy to prevent fuel clad rupture, plus provisions to handle short cooled fuel and criteria to ensure safety during handling. In addition the handling and elimination of residual sodium must be

  10. Development of a thermal–hydraulic analysis code for the Pebble Bed Water-cooled Reactor

    International Nuclear Information System (INIS)

    Highlights: ► Main design features of the PBWR were put forward. ► Thermal–hydraullics analysis code for the PBWR was developed and verified. ► Key thermal–hydraullics parameters were calculated in normal operation. ► The PBWR has a great pressure loss but an excellent heat transfer characteristic. ► Maximum fuel temperature and MDNBR are in conformity with safety criterion. - Abstract: The Pebble Bed Water-cooled Reactor (PBWR) is a water-moderated water-cooled pebble bed reactor in which millions of tristructural-isotropic (TRISO) coated micro-fuel elements (MFE) pile in each assembly. Light water is used as coolant that flows from bottom to top in the assembly while the moderator water flows in the reverse direction out of the assembly. Steady-state thermal–hydraullic analysis code for the PBWR will provide a set of thermal hydraulic parameters of the primary loop so that heat transported out of the core can match with the heat generated by the core for a safe operation of the reactor. The key parameters of the core including the void fraction, pressure drop, heat transfer coefficients, the temperature distribution and the Departure from Nucleate Boiling Ratio (DNBR) is calculated for the core in normal operation. The code can calculate for liquid region, water-steam two phase region and superheated steam region. The results show that the maximum fuel temperature is much lower than the design limitation and the flow distribution can meet the cooling requirement in the reactor core. As a new type of nuclear reactor, the main design features with a sufficient safety margin were also put forward in this paper.

  11. The Addition of Noncondensable Gases into RELAP5-3D for Analysis of High Temperature Gas-Cooled Reactors

    International Nuclear Information System (INIS)

    Oxygen, carbon dioxide, and carbon monoxide have been added to the RELAP5-3D computer code as noncondensable gases to support analysis of high temperature gas-cooled reactors. Models of these gases are required to simulate the effects of air ingress on graphite oxidation following a loss-of-coolant accident. Correlations were developed for specific internal energy, thermal conductivity, and viscosity for each gas at temperatures up to 3000 K. The existing model for internal energy (a quadratic function of temperature) was not sufficiently accurate at these high temperatures and was replaced by a more general, fourth-order polynomial. The maximum deviation between the correlations and the underlying data was 2.2% for the specific internal energy and 7% for the specific heat capacity at constant volume. The maximum deviation in the transport properties was 4% for oxygen and carbon monoxide and 12% for carbon dioxide

  12. 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.

  13. 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.

  14. 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

  15. Improving Fuel Cycle Design and Safety Characteristics of a Gas Cooled Fast Reactor

    International Nuclear Information System (INIS)

    The Gas Cooled Fast Reactor (GCFR)is one of the Generation IV reactor concepts. This concept specifically targets sustainability of nuclear power generation. In nuclear reactors fertile material is converted to fissile fuel. If the neutrons inducing fission are highly energetic, the opportunity exists to convert more than one fertile nucleus per fission, thereby effectively breeding new nuclear fuel. Reactors operating on this principle are called ‘Fast Breeder Reactor’. Since natural uranium contains 99.3%of the fertile isotope 238U, breeding increases the energy harvested from the nuclear fuel. If nuclear energy is to play an important role as a source of energy in the future, fast breeder reactors are essential for breeding nuclear fuel. Fast neutrons are also more efficient to destruct heavy (Minor Actinide, MA) isotopes, such as Np, Am and Cm isotopes, which dominate the long-term radioactivity of nuclear waste. So the waste life-time can be shortened if the MA nuclei are destroyed. An important prerequisite of sustainable nuclear energy is the closed fuel cycle, where only fission products are discharged to a final repository, and all Heavy Metal (HM) are recycled. The reactor should breed just enough fissile material to allow refueling of the same reactor, adding only fertile material to the recycled material. Other key design choices are highly efficient power conversion using a direct cycle gas turbine, and better safety through the use of helium, a chemically inert coolant which cannot have phase changes in the reactor core. Because the envisaged core temperatures and operating conditions are similar to thermal-spectrum High Temperature Reactor (HTR) concepts, the research for this thesis initially focused on a design based on existing HTR fuel technology: coated particle fuel, assembled into fuel assemblies. It was found that such a fuel concept could not meet the Generation IV criteria set for GCFR: self-breeding is difficult, the temperature

  16. 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.

  17. 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.

  18. Seismic isolation of lead-cooled reactors: The European project SILER

    Energy Technology Data Exchange (ETDEWEB)

    Forni, Massimo; Poggianti, Alessandro; Scipinotti, Riccardo [ENEA, Via Martiri di Monte Sole, Bologna (Italy); Dusi, Alberto; Manzoni, Elena [Numeria Engineering srl, Galleria del Corso, Cremona (Italy)

    2014-10-15

    SILER (Seismic-Initiated event risk mitigation in LEad-cooled Reactors) is a Collaborative Project, partially funded by the European Commission in the 7th Framework Programme, aimed at studying the risk associated to seismic-initiated events in Generation IV Heavy Liquid Metal reactors, and developing adequate protection measures. The project started in October 2011, and will run for a duration of three years. The attention of SILER is focused on the evaluation of the effects of earthquakes, with particular regards to beyond-design seismic events, and to the identification of mitigation strategies, acting both on structures and components design. Special efforts are devoted to the development of seismic isolation devices and related interface components. Two reference designs, at the state of development available at the beginning of the project and coming from the 6th Programme, have been considered: ELSY (European Lead Fast Reactor) for the Lead Fast Reactors (LFR), and MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) for the Accelerator-Driven Systems (ADS). This paper describes the main activities and results obtained so far, paying particular attention to the development of seismic isolators, and the interface components which must be installed between the isolated reactor building and the non-isolated parts of the plant, such as the pipe expansion joints and the joint-cover of the seismic gap.

  19. 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.

  20. Educational laboratory based on a multifunctional analyzer of a reactor of a nuclear power plant with a water-moderated water-cooled reactor

    International Nuclear Information System (INIS)

    Authors presents an educational laboratory Safety and Control of a Nuclear Power Facility established by the Department of Automation for students and specialists of the nuclear power industry in the field of control, protection, and safe exploitation of reactor facilities at operating, constructing, and designing nuclear power plants with water-moderated water-cooled reactors

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

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

    International Nuclear Information System (INIS)

    For nuclear power plant application, applying S-CO2 Brayton cycle to Sodium cooled Fast Reactors and Small Modular Reactors are currently considered and active research is being performed by various research institutions and universities. As a part of research activities on the SCO2 Brayton cycle development for a nuclear power system, KAIST joint research team is currently working on an innovative Sodium cooled Fast Reactor (iSFR) development which utilizes S-CO2 Brayton cycle as its power conversion system. Various research subjects including reactor physics, thermo-hydraulics, material, cycle analysis and system integration are being considered as research issues currently. However, technical issues rising from dramatic change of thermodynamic property of CO2 near the critical point still remain as problems to be solved. As a result, 3D impeller model generation based on 1D mean stream line analysis results was successfully performed for non-airfoil blades. Since 3D model generation module works successfully, KAISTTMD can support 3D CFD analysis for internal flow structure in the designed impeller. Compressor loss mechanisms are complex phenomena and these are difficulties to be modeled while considering each loss mechanism separately

  3. 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.

  4. 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

  5. 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.

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

    International Nuclear Information System (INIS)

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

  7. 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).

  8. 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

  9. The lead cooled fast reactor benchmark Brest-300: analysis with sensitivity method

    International Nuclear Information System (INIS)

    Lead cooled fast neutrons reactor is one of the most interesting candidates for the development of atomic energy. BREST-300 is a 300 MWe lead cooled fast reactor developed by the NIKIET (Russia) with a deterministic safety approach which aims to exclude reactivity margins greater than the delayed neutron fraction. The development of innovative reactors (lead coolant, nitride fuel...) and fuel cycles with new constraints such as cycle closure or actinide burning, requires new technologies and new nuclear data. In this connection, the tool and neutron data used for the calculational analysis of reactor characteristics requires thorough validation. NIKIET developed a reactor benchmark fitting of design type calculational tools (including neutron data). In the frame of technical exchanges between NIKIET and EDF (France), results of this benchmark calculation concerning the principal parameters of fuel evolution and safety parameters has been inter-compared, in order to estimate the uncertainties and validate the codes for calculations of this new kind of reactors. Different codes and cross-sections data have been used, and sensitivity studies have been performed to understand and quantify the uncertainties sources.The comparison of results shows that the difference on keff value between ERANOS code with ERALIB1 library and the reference is of the same order of magnitude than the delayed neutron fraction. On the other hand, the discrepancy is more than twice bigger if JEF2.2 library is used with ERANOS. Analysis of discrepancies in calculation results reveals that the main effect is provided by the difference of nuclear data, namely U238, Pu239 fission and capture cross sections and lead inelastic cross sections

  10. Comparative Analysis of Effectiveness of Various Emergency Core Cooling System Design Options for Sodium Fast Reactors of High Rower

    International Nuclear Information System (INIS)

    Effectiveness of various design options for emergency core cooling systems has been compared as applied to a pool type sodium fast reactor of high power. Thermal hydraulic parameters of the reactor under cooling conditions are analyzed with the use of the Russian thermal hydraulic code GRIF which allows 3D velocity and temperature fields to be calculated in the reactor, with account of thermal hydraulic processes in the core inter-wrapper space. To realize the cooling system margin in case of additional parallel failures, the parameters were calculated for the cooling mode accompanied with additional conditions of malfunction of part of emergency heat exchangers (DHX) for unknown reasons. Based one the calculation analysis the conclusion is made about a relative effectiveness of the emergency cooling system design options considered. (author)

  11. 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.

  12. 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...

  13. 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.).

  14. 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.

  15. Comparison of different air supply and cooling systems in pig fattening houses; Vergleich von Zuluftfuehrungs- und Kuehlungssystemen in der Schweinemast

    Energy Technology Data Exchange (ETDEWEB)

    Threm, Joachim; Gallmann, Eva; Jungbluth, Thomas [Hohenheim Univ. (Germany). Fachgebiet Verfahrenstechnik der Tierhaltungssysteme; Pflanz, Wilhelm [Bildungs- und Wissenszentrum LSZ, Boxberg (Germany). Referat Haltungssysteme, Stallbau, Stallklima, Biogas

    2011-07-01

    There is a need for research regarding the optimization of air supply and of air cooling systems of livestock houses for fatteners. A decision support project for the German Ministry of Agriculture has been established to investigate such technologies on a research farm and under practical conditions. Since April 2011, on the research farm Landesanstalt fuer Schweinezucht (LSZ Boxberg) measurements on three air supply and cooling system are carried out: Underfloor air inlet and supply, cooling pad and high pressure evaporative indoor air cooling. First results show differences concerning the parameters temperature, differential pressure and air velocity. (orig.)

  16. Direct-Drive Gas-Cooled Reactor Power System: Concept and Preliminary Testing

    Science.gov (United States)

    Wright, S. A.; Lipinski, R. J.; Godfroy, T. J.; Bragg-Sitton, S. M.; VanDyke, M. K.

    2002-01-01

    This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet- sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrically heated testing of simulated reactor components.

  17. 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.

  18. Uncertainty analysis of infinite homogeneous lead and sodium cooled fast reactors at beginning of life

    International Nuclear Information System (INIS)

    The objective of the present work is to estimate breeding ratio, radiation damage rate and minor actinide transmutation rate of infinite homogeneous lead and sodium cooled fast reactors. Uncertainty analysis is performed taking into account uncertainty in nuclear data and composition of the reactors. We use the recently released ENDF/B-VII.1 nuclear data library and restrict the work to the beginning of reactor life. We work under multigroup approximation. The Bondarenko method is used to acquire effective cross sections for the homogeneous reactor. Modeling error and numerical error are estimated. The adjoint sensitivity analysis is performed to calculate generalized adjoint fluxes for the responses. The generalized adjoint fluxes are used to calculate first order sensitivities of the responses to model parameters. The acquired sensitivities are used to propagate uncertainties in the input data to find out uncertainties in the responses. We show that the uncertainty in model parameters is the dominant source of uncertainty, followed by modeling error, input data precision and numerical error. The uncertainty due to composition of the reactor is low. We identify main sources of uncertainty and note that the low-fidelity evaluation of 16O is problematic due to lack of correlation between total and elastic reactions

  19. 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.

  20. 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.

  1. Aspects of Safety Analysis for Sodium Cooled Fast Reactor Design and Licensing

    International Nuclear Information System (INIS)

    This paper discusses aspects of safety analysis of a sodium cooled fast reactor (SFR) that are necessary for nuclear plant design and licensing. In the last two decades, SFR safety analysis has focused mostly on anticipated transients without scram (or unprotected transients). These include unprotected loss of flow, unprotected transient over power and unprotected loss of heat sink. While these transients provide a bounding safety assessment, other more probable transients need to be assessed to support SFR design and licensing efforts. Using a representative traveling wave reactor (TWR) reactor core, different aspects of safety analysis are discussed with example results from SASSYS-1/SAS4A calculations and their implications on SFR design. First, transient event classification is presented. Then, the effects of protected transients (where the reactor scrams) on various design parameters (scram set-points, delay time, peak cladding temperature limit etc.) are demonstrated. Third, the effects of uncertainties of various parameters are demonstrated. Finally, sensitivity analyses and their implications on the reactor design and safety profiles are discussed. (author)

  2. 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.

  3. Cooling energy efficiency and classroom air environment of a school building operated by the heat recovery air conditioning unit

    International Nuclear Information System (INIS)

    The recently-built school buildings have adopted novel heat recovery ventilator and air conditioning system. Heat recovery efficiency of the heat recovery facility and energy conservation ratio of the air conditioning unit were analytically modeled, taking the ventilation networks into account. Following that, school classroom displacement ventilation and its thermal stratification and indoor air quality indicated by the CO2 concentration have been numerically modeled concerning the effects of delivering ventilation flow rate and supplying air temperature. Numerical results indicate that the promotion of mechanical ventilation rate can simultaneously boost the dilution of indoor air pollutants and the non-uniformity of indoor thermal and pollutant distributions. Subsequent energy performance analysis demonstrates that classroom energy demands for ventilation and cooling could be reduced with the promotion of heat recovery efficiency of the ventilation facility, and the energy conservation ratio of the air conditioning unit decreases with the increasing temperatures of supplying air. Fitting correlations of heat recovery ventilation and cooling energy conservation have been presented. - Highlights: • Low energy school buildings and classroom environment. • Heat recovery facility operating with an air conditioning unit. • Displacement ventilation influenced by the heat recovery efficiency. • Energy conservation of cooling and ventilation through heat recovery. • Enhancement of classroom environment with reduction of school building energy

  4. Efficiency Testing of the Air Cleaning System for a High Temperature Reactor

    International Nuclear Information System (INIS)

    The Los Alamos Ultra High Temperature Reactor Experiment (UHTREX) utilizes a helium-cooled, graphite-moderated reactor, employing refractory fuel elements. Under accident conditions, the effluent that may be released from this reactor requires an air-cleaning system capable of reducing radioactive gas and particulate contaminants to safe levels. Dioctyl phthalate and iodine-131 were used as test aerosols for the HEPA and activated carbon filters, respectively. Methods of aerosol generation and test procedures are detailed for the preinstallation tests of the carbon and in-place testing of the carbon and HEPA filters. The importance of visual inspection of the HEPA filters prior to installation and supervision of filter installation is discussed. In-place tests indicated desirable design changes which would (1) simplify in-place testing procedures, (2) expedite installation and future changing of the filters, and (3) ensure operation of a more efficient system. Problems encountered during in-place testing, recommendations for the design of similar systems, and acceptance criteria used at LASL are discussed. (author)

  5. 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.

  6. Safety analysis code 'COOLTMP' for assessment of PHT cooling under reactor shutdown conditions

    International Nuclear Information System (INIS)

    The thermal energy generated by the reactor core is removed by the Primary Heat Transport (PHT) System when the reactor is under normal operation, by operation of the primary circulation pumps and steam generators. However, when the reactor is shutdown, the decay heat removal is done by the Shut Down (S/D) Cooling heat exchangers and pumps of lower capacity. In the event of loss/stoppage of circulation of PHT under such a situation, the bulk of the decay heat generated will be distributed to the moderator system, end shield system and through the feeders to the feeder cabinet/FM vault environment. However, the PHT inventory in the channel will be heated up because of loss of flow in the channel. The code COOLTMP has been developed to estimate the temperature of PHT following a loss/stoppage of circulation, when the reactor is under shutdown condition. It predicts the increase in the PHT temperature with time for hot channel, average channel or a specific channel under such a condition. It also calculates the apportionment of the decay heat to different heat sinks, viz. moderator, end shield and FM Vault. This computation is required when the plant is required to be under shutdown for doing some maintenance job on the PHT system, feeders or channels where the S/D cooling system has to be stopped and in some cases the headers have to be drained. At that time such a calculation gives whether the peak PHT temperature, or the time available to reach such a temperature, as obtained, is acceptable to carry out such a job. Hence, the schedule of the maintenance job can be decided. This code has been validated for RAPS and MAPS and used extensively for predicting PHT temperature after reactor shutdown to obtain regulatory clearances to stop forced circulation with and without header filled. (author)

  7. 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.

  8. Development of fluorocarbon rubber for backup seals of sodium cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Highlights: → Negligible chemical degradation of seal compound during ageing (in unstrained state) in air at 140/170/200 oC for 32 weeks. → Cross-link exchange, Joule-Gough effect and ionic interaction during ageing in unstrained state. → Enhanced physical/chemical degradation of compound during ageing under strain. → Capability of compound to withstand heat, radiation, air and mechanical load in reactor for 10 years. → Negligible chemical dose rate effect and gas evolution from compound during seal operation. -- Abstract: The development of a fluorohydrocarbon rubber compound for static backup seals of 500 MWe, Prototype Fast Breeder Reactor (PFBR) is depicted. Variations of a previously developed Viton A-401C based formulation were subjected to processability tests, accelerated heat ageing in air, mechanical characterization and production trials. Finite element analysis and literature data extrapolation were combined with long term ageing to ascertain the life (minimum 10 years) of chosen formulation in reactor under synergistic influences of 110 oC, 23 mGy/h (γ dose rate) and air considering postulated accidental conditions. Validation of test seals and quality assessment indicate that composition and properties of the validated laboratory compound has been translated effectively to the reactor seals, installed recently in PFBR. The tensile and hardness specimens indicated negligible degradation and exceptional thermo-oxidative stability of the seal compound during ageing (32 weeks at 140/170/200 oC) even though interesting manifestations of cross-link exchange and ionic interactions were observed. Compression set results, showing definite trends of change under ageing and stain, were used in Arrhenius and Williams Landel Ferry equations for realistic life prediction. The development provides a foundation to simplify and standardize the design, development and operation of major elastomeric sealing applications of Indian nuclear reactors based on a

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

    International Nuclear Information System (INIS)

    Highlights: • Thorium as support fertile material for TRU transmutation in Fast Reactors. • Comparative analysis of Th and U based breakeven and burner Fast Reactors. • Thorium fosters significant advantages in terms of safety parameters. • Inherent safety is investigated through quasi-static reactivity and energy balances. • Th use in low-CR Fast Reactors does not reduce fuel decay heat and neutron sources. - Abstract: 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

  10. 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.

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

    International Nuclear Information System (INIS)

    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

  12. Performance studies of a new core cooling monitor in a boiling water reactor

    International Nuclear Information System (INIS)

    Performance studies of a new type of core cooling monitors have been carried out in the Barsebaeck Nuclear Power Station during the operation periods 1988-10-04 to 1989-07-05, 1989-08-03 to 1990-09-05 and 1990-09-28 to 1991-07-04. The results showed that the monitors, which were placed inside the reactor core, are very sensitive to variations of the reactor operating conditions, and that 34 months of irradiation did not influence the signals from the monitors. Experiments were also carried out in a 160 bar loop, where sudden uncovers of the monitors were achieved by decreasing the liquid level of the coolant surrounding the monitors. The experiments included the pressures of 5, 20, 50, 70 and 155 bar, and the responses to uncover were in the ranges between 11 and 82 mV/sec or a total step change of 2 V at typical BWR conditions. This is of the order of two decades higher than the responses from monitors based on thermocouple readings. The monitors can be operated in two modes, the core cooling mode and the temperature mode. In the former mode the electrical current is 3-4 A, and in the latter mode, where the monitor actually serves as a thermometer, the current is in the order of 50-100 mA. In the laboratory the monitors have been studied for temperatures up to 1265 deg. C, which is very useful in case of a severe reactor accident. Thus, during such events the temperatures in the reactor core could be followed up to this level and the monitors could also be used to activate certain safety equipment. The function as well as the design of the instrument is verified in laboratory experiments, computer calculations and reactor tests and is now ready for implementation in the BWR instrumentation. In summary: 1. The proposed monitor can operate in two modes; the core cooling mode and the temperature mode. 2. Laboratory studies have shown that the responses to uncover are two decades higher than signals from monitors based on thermocouple readings. 3. No effects of

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

    International Nuclear Information System (INIS)

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

  14. 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.

  15. Basic study on high temperature gas cooled reactor technology for hydrogen production

    International Nuclear Information System (INIS)

    The annual production of hydrogen in the world is about 500 billion m3. Currently hydrogen is consumed mainly in chemical industries. However hydrogen has huge potential to be consumed in transportation sector in coming decades. Assuming that 10% of fossil energy in transportation sector is substituted by hydrogen in 2020, the hydrogen in the sector will exceed current hydrogen consumption by more than 2.5 times. Currently hydrogen is mainly produced by steam reforming of natural gas. Steam reforming process is chiefest way to produce hydrogen for mass production. In the future, hydrogen has to be produced in a way to minimize CO2 emission during its production process as well as to satisfy economic competition. One of the alternatives to produce hydrogen under such criteria is using heat source of high-temperature gas-cooled reactor. The high-temperature gas-cooled reactor represents one type of the next generation of nuclear reactors for safe and reliable operation as well as for efficient and economic generation of energy

  16. Draft of standard for graphite core components in high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    For the design of the graphite components in the High Temperature Engineering Test Reactor (HTTR), the graphite structural design code for the HTTR etc. were applied. However, general standard systems for the High Temperature Gas-cooled Reactor (HTGR) have not been established yet. The authors had studied on the technical issues which is necessary for the establishment of a general standard system for the graphite components in the HTGR. The results of the study were documented and discussed at a 'Special committee on research on preparation for codes for graphite components in HTGR' at Atomic Energy Society of Japan (AESJ). As a result, 'Draft of Standard for Graphite Core Components in High Temperature Gas-cooled Reactor.' was established. In the draft standard, the graphite components are classified three categories (A, B and C) in the standpoints of safety functions and possibility of replacement. For the components in the each class, design standard, material and product standards, and in-service inspection and maintenance standard are determined. As an appendix of the design standard, the graphical expressions of material property data of 1G-110 graphite as a function of fast neutron fluence are expressed. The graphical expressions were determined through the interpolation and extrapolation of the irradiated data. (author)

  17. Measurement of sulphur-35 in the coolant gas of the Windscale Advanced Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Sulphur is an important element in some food chains and the release of radioactive sulphur to the environment must be closely controlled if the chemical form is such that it is available or potentially available for entering food chains. The presence of sulphur-35 in the coolant gas of the Windscale Advanced Gas-Cooled Reactor warranted a study to assess the quantity and chemical form of the radioactive sulphur in order to estimate the magnitude of the potential environmental hazard which might arise from the release of coolant gas from Civil Advanced Gas-Cooled Reactors. A combination of gas chromatographic and radiochemical analyses revealed carbonyl sulphide to be the only sulphur-35 compound present in the coolant gas of the Windscale Reactor. The concentration of carbonyl sulphide was found to lie in the range 40 to 100 x 10-9 parts by volume and the sulphur-35 specific activity was about 20 mCi per gramme. The analytical techniques are described in detail. The sulphur-35 appears to be derived from the sulphur and chlorine impurities in the graphite. A method for the preparation of carbonyl sulphide labelled with sulphur-35 is described. (author)

  18. 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.

  19. The role of the IAEA in gas-cooled reactor development and application

    International Nuclear Information System (INIS)

    Within the Statute establishing the International Atomic Energy Agency there are several functions authorized for the Agency. One of these functions is ''to encourage and assist research on, and development and practical application of, atomic energy for peaceful uses throughout the world...''. The development of nuclear power is deemed an important application of this function. The representatives of Member States with national gas cooled reactor (GCR) programmes advise the Agency on its activities in the development and application of the GCR. The committee of leaders in GCR technology representing these Member States is the International Working Group on Gas Cooled Reactors (IWGGCR). The activities carried out by the Agency under the frame of the IWGGCR include technical information exchange meetings and cooperative Coordinated Research Programmes. Within the technical information exchange meetings are Specialist Meetings to review progress on selected technology areas and Technical Committee Meetings and Workshops for more general participation. Consultancies and Advisory Group Meetings are convened to provide the Agency with advise on specific technical matters. The Coordinated Research Programmes (CRPs) established within the frame of the IWGGCR for the GCR programme include: Validation of Safety Related Physics Calculations for Low Enriched GCRs; Validation of Predictive Methods for Fuel and Fission Product Behaviour in GCRs; Heat Transport and Afterheat Heat Removal for GCRs under Accident Conditions; and Design and Evaluation of Heat Utilization Systems for the High Temperature Engineering Test Reactor. This paper summarizes the role of the International Atomic Energy Agency in GCR technology development and application. (author). 6 refs, 3 tabs

  20. Light water cooled, high temperature and high performance nuclear power plants concept of once-through coolant cycle, supercritical-pressure, light water cooled nuclear reactors

    International Nuclear Information System (INIS)

    Supercritical-pressure, light water cooled nuclear reactors corresponding to nuclear reactors of once-through boilers, are of theoretical development from LWR. Under supercritical pressure, a steam turbine can be driven directly with cooled water with high enthalpy, as not seen boiling and required for recycling. The reactor has no steam-water separation and recycling systems on comparison with the boiling water type LWR, and is the same once-through type as supercritical-pressure thermal power generation plants. Then, all of cooling water at reactor core are sent to turbine. The reactor has no steam generator, and pressurizer, on comparison with PWR. As it requires no steam-water separator, steam drier, and recycling system on comparison with BWR, it becomes of smaller size and has shape and size nearly equal to those of PWR. And, its control bars can be inserted from upper direction like PWR, and can use its driving system. Here was introduced some concepts on high-temperature and high-performance light water reactor, nuclear power generation using a technology on supercritical-pressure thermal power generation. (G.K.)

  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. 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.

  3. Technologies for gas cooled reactor decommissioning, fuel storage and waste disposal. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    Gas cooled reactors (GCRs) and other graphite moderated reactors have been important part of the world's nuclear programme for the past four decades. The wide diversity in status of this very wide spectrum of plants from initial design to decommissioning was a major consideration of the International Working group on Gas Cooled Reactors which recommended IAEA to convene a Technical Committee Meeting dealing with GCR decommissioning, including spent fuel storage and radiological waste disposal. This Proceedings includes papers 25 papers presented at the Meeting in three sessions entitled: Status of Plant Decommissioning Programmes; Fuels Storage Status and Programmes; waste Disposal and decontamination Practices. Each paper is described here by a separate abstract

  4. 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.

  5. 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

    compared to without cooling. The acceptability of the thermal environment was similar for all cooling devices. The acceptability of air movement and PAQ increased when the local cooling methods were used. The best results were achieved with personalized ventilation and cooling fan. The improvement in PAQ...

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

    Science.gov (United States)

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

    1975-01-01

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

  7. Power boost of gas turbines by inlet air cooling

    Energy Technology Data Exchange (ETDEWEB)

    White, C.; Raghu, S. [State Univ. of New York, Stony Brook, NY (United States). Dept. of Mechanical Engineering; Giannotti, G.; Giannotti, H. [Giannotti Associates, Bellport, NY (United States)

    1996-12-31

    The design of a power boost system for an existing gas turbine unit using a direct spray evaporative spray cooling method is described in this paper. Experiments were conducted to determine the extent of cooling possible by this method, droplet size requirements and spray water requirements. Although up to 4.9% of power boost is theoretically possible at the design conditions of 32.2 C (90 F) and 60% relative humidity, various other constraints limit the actual power boost to about 3.8% of the nominal power. For small values of the wet-bulb depression (less than about 5 C or 10 F), the required droplet sizes for complete evaporation become so small (of the order of 5--10 microns with high flow rates) that production of them at the required flow rates is a challenging task.

  8. Mitigation of sodium-cooled fast reactor severe accident consequences using inherent safety principles

    International Nuclear Information System (INIS)

    Full text: Sodium-cooled fast reactors are designed to have a high level of safety. Events of high probability of occurrence are typically handled without consequence through reliable engineering systems and good design practices. For accidents of lower probability, the initiating events are characterized by larger and more numerous challenges to the reactor system, such as failure of one or more major engineered systems and can also include a failure to scram the reactor in response. As the initiating conditions become more severe, they have the potential for creating serious consequences of potential safety significance, including fuel melting, fuel pin disruption and recriticality. If the progression of such accidents is not mitigated by design features of the reactor, energetic events and dispersal of radioactive materials may result. In the United States, accidents which have the potential for severe consequences usually are of probability less than 1 x 10-4 per reactor year, intended to satisfy the U.S. Nuclear Regulatory Commission (NRC) goal of limiting accidents with any fuel melting to such low probabilities. Such severe accidents include the category of Anticipated Transient Without Scram (ATWS) events mentioned above. Three accidents are usually analyzed to evaluate the reactor response in these cases; the unprotected (unscrammed) loss-of-flow (ULOF), where pumping power is lost and the pumps coast down, reducing coolant flow through the reactor core; the unprotected transient overpower (UTOP), where a control rod is inadvertently withdrawn from the core; and the unprotected loss-of-heat-sink (ULOHS), where the steam generator is isolated from the reactor in response to a turbine trip. For each of these accidents, there are several approaches that can be used to mitigate the consequences of such severe accident initiators, which typically include fuel pin failures and core disruption. One approach is to increase the reliability of the reactor protection

  9. Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs)

    International Nuclear Information System (INIS)

    The supercritical water cooled reactor (SCWR) is an innovative water cooled reactor concept which uses water pressurized above its thermodynamic critical pressure as the reactor coolant. This concept offers high thermal efficiencies and a simplified reactor system, and is hence expected to help to improve economic competitiveness. Various kinds of SCWR concepts have been developed, with varying combinations of reactor type (pressure vessel or pressure tube) and core spectrum (thermal, fast or mixed). There is great interest in both developing and developed countries in the research and development (R&D) and conceptual design of SCWRs. Considering the high interest shown in a number of Member States, the IAEA established in 2008 the Coordinated Research Project (CRP) on Heat Transfer Behaviour and Thermo-hydraulics Code Testing for SCWRs. The aim was to foster international collaboration in the R&D of SCWRs in support of Member States’ efforts and under the auspices of the IAEA Nuclear Energy Department’s Technical Working Groups on Advanced Technologies for Light Water Reactors (TWG-LWR) and Heavy Water Reactors (TWG-HWR). The two key objectives of the CRP were to establish accurate databases on the thermohydraulics of supercritical pressure fluids and to test analysis methods for SCWR thermohydraulic behaviour to identify code development needs. In total, 16 institutes from nine Member States and two international organizations were involved in the CRP. The thermohydraulics phenomena investigated in the CRP included heat transfer and pressure loss characteristics of supercritical pressure fluids, development of new heat transfer prediction methods, critical flow during depressurization from supercritical conditions, flow stability and natural circulation in supercritical pressure systems. Two code testing benchmark exercises were performed for steady state heat transfer and flow stability in a heated channel. The CRP was completed with the planned outputs in

  10. Performance and economic enhancement of cogeneration gas turbines through compressor inlet air cooling

    Science.gov (United States)

    Delucia, M.; Bronconi, R.; Carnevale, E.

    1994-04-01

    Gas turbine air cooling systems serve to raise performance to peak power levels during the hot months when high atmospheric temperatures cause reductions in net power output. This work describes the technical and economic advantages of providing a compressor inlet air cooling system to increase the gas turbine's power rating and reduce its heat rate. The pros and cons of state-of-the-art cooling technologies, i.e., absorption and compression refrigeration, with and without thermal energy storage, were examined in order to select the most suitable cooling solution. Heavy-duty gas turbine cogeneration systems with and without absorption units were modeled, as well as various industrial sectors, i.e., paper and pulp, pharmaceuticals, food processing, textiles, tanning, and building materials. The ambient temperature variations were modeled so the effects of climate could be accounted for in the simulation. The results validated the advantages of gas turbine cogeneration with absorption air cooling as compared to other systems without air cooling.

  11. Overview of direct air free cooling and thermal energy storage potential energy savings in data centres

    International Nuclear Information System (INIS)

    In the last years the total energy demand of data centres has experienced a dramatic increase which is expected to continue. This is why data centres industry and researchers are working on implementing energy efficiency measures and integrating renewable energy to overcome energy dependence and to reduce operational costs and CO2 emissions. The cooling system of these unique infrastructures can account for 40% of the total energy consumption. To reduce the energy consumption, free cooling strategies are used more and more, but so far there has been little research about the potential of thermal energy storage (TES) solutions to match energy demand and energy availability. Hence, this work intends to provide an overview of the potential of the integration of direct air free cooling strategy and TES systems into data centres located at different European locations. For each location, the benefit of using direct air free cooling is evaluated energetically and economically for a data centre of 1250 kW. The use of direct air free cooling is shown to be feasible. This does not apply the TES systems by itself. But when using TES in combination with an off-peak electricity tariff the operational cooling cost can be drastically reduced. - Highlights: • The total annual hours for direct air free cooling in data centres are calculated. • The potential of TES integration in data centres is evaluated. • The implementation of TES to store the ambient air cold is not recommended. • TES is feasible if combined with redundant chillers and off-peak electricity price. • The cooling electricity cost is being reduced up to 51%, depending on the location

  12. High temperature gas cooled reactor technology development. Proceedings of a technical committee meeting

    International Nuclear Information System (INIS)

    The successful introduction of an advanced nuclear power plant programme depends on many key elements. It must be economically competitive with alternative sources of energy, its technical development must assure operational dependability, the support of society requires that it be safe and environmentally acceptable, and it must meet the regulatory standards developed for its use and application. These factors interrelate with each other, and the ability to satisfy the established goals and criteria of all of these requirements is mandatory if a country or a specific industry is to proceed with a new, advanced nuclear power system. It was with the focus on commercializing the high temperature gas cooled reactor (HTGR) that the IAEA's International Working Group on Gas Cooled Reactors recommended this Technical Committee Meeting (TCM) on HTGR Technology Development. Over the past few years, many Member States have instituted a re-examination of their nuclear power policies and programmes. It has become evident that the only realistic way to introduce an advanced nuclear power programme in today's world is through international co-operation between countries. The sharing of expertise and technical facilities for the common development of the HTGR is the goal of the Member States comprising the IAEA's International Working Group on Gas Cooled Reactors. This meeting brought together key representatives and experts on the HTGR from the national organizations and industries of ten countries and the European Commission. The state electric utility of South Africa, Eskom, hosted this TCM in Johannesburg, from 13 to 15 November 1996. This TCM provided the opportunity to review the status of HTGR design and development activities, and especially to identify international co-operation which could be utilized to bring about the commercialization of the HTGR

  13. KUEBEL. A Fortran program for computation of cooling-agent-distribution within reactor fuel-elements

    International Nuclear Information System (INIS)

    KUEBEL is a Fortran-program for computation of cooling-agent-distribution within reactor fuel-elements or -zones of theirs. They may be assembled of max. 40 cooling-channels with laminar up to turbulent type of flow (respecting Reynolds' coefficients up to 2.0E+06) at equal pressure loss. Flow-velocity, dynamic flow-, contraction- and friction-losses will be calculated for each channel and for the total zone. Other computations will present mean heat-up of cooling-agent, mean outlet-temperature of the core, boiling-temperature and absolute pressure at flow-outlet. All characteristic coolant-values, including the factor of safety for flow-instability of the most-loaded cooling gap are computed by 'KUEBEL' too. Absolute pressure at flow-outlet or is-factor may be defined as dependent or independent variables of the program alternatively. In latter case 3 variations of solution will be available: Adapted flow of cooling-agent, inlet-temperature of the core and thermal power. All calculations can be done alternatively with variation of parameters: flow of cooling-agent, inlet-temperature of the core and thermal power, which are managed by the program itself. 'KUEBEL' is able to distinguish light- and heavy-water coolant, flow-direction of coolant and fuel elements with parallel, rectangular, respectively concentric, cylindrical shape of their gaps. Required material specifics are generated by the program. Segments of fuel elements or constructively unconnected gaps can also be computed by means of interposition of S.C. 'phantom channels'. (orig.)

  14. External Reactor Vessel Cooling Strategy Application for APR1400 Station Blackout Scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Tran, Quang Diep Khanh; Oh, Seung Jong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of); Ro, Seo Mi [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    In Korea, all NPPs implemented post-Fukushima measures. In U. S., the approach is to expand B5b strategy to build up FLEX strategies utilizing portable equipment. With portable pumps and connections, one can mitigate and respond to wide range of accidents. In this study, we examined the feasibility of cavity flooding using portable pumps for ERVC and its effectiveness in responding to extended SBO scenarios. The external reactor vessel cooling (ERVC) is one of important severe accident management (SAM) strategies adopted for APR1400. ERVC strategy is to remove decay heat from molten corium in lower plenum by submerging the vessel with water. The strategy requires water injection into cavity and APR1400 uses shutdown cooling pumps for this purpose, requiring electricity. Recent Fukushima accident shows the importance of mitigation capability against extended SBO scenarios.

  15. Evaluation of sprayed chromium carbide coatings for gas-cooled reactor applications

    International Nuclear Information System (INIS)

    Sprayed chromium carbide-nichrome coatings are candidates for protection of faying and sliding surfaces of critical components of gas-cooled reactors from friction and wear damage. These coatings must provide protection throughout the reactor lifetime under high temperature exposure conditions. Extensive evaluation work to characterize these coatings is underway. The work includes studies of friction and wear behavior in helium; stability of the coatings in a low oxygen potential helium environment; impure helium corrosion of coated specimens; and the effect of the coatings on mechanical properties of the substrate alloy. Much of the work reported is on the evaluation of plasma-sprayed coatings. However, a brief discussion of the behavior of coatings applied by the detonation-gun process and high-energy plasma-gun processes is also included

  16. 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.

  17. Modular High Temperature Gas-Cooled Reactor heat source for coal conversion

    International Nuclear Information System (INIS)

    In the industrial nations, transportable fuels in the form of natural gas and petroleum derivatives constitute a primary energy source nearly equivalent to that consumed for generating electric power. Nations with large coal deposits have the option of coal conversion to meet their transportable fuel demands. But these processes themselves consume huge amounts of energy and produce undesirable combustion by-products. Therefore, this represents a major opportunity to apply nuclear energy for both the environmental and energy conservation reasons. Because the most desirable coal conversion processes take place at 800 degree C or higher, only the High Temperature Gas-Cooled Reactors (HTGRs) have the potential to be adapted to coal conversion processes. This report provides a discussion of this utilization of HTGR reactors

  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. Use of a supercritical water-cooled reactor for process heat to support thermochemical hydrogen production

    International Nuclear Information System (INIS)

    The SuperCritical Water-cooled nuclear Reactor (SCWR) is one of six Generation-IV nuclear-reactor concepts currently under development worldwide. It is designed to operate at pressures of 25 MPa and temperatures up to 625°C. These operating conditions make an SCW Nuclear Power Plant (NPP) suitable to support thermochemical-based hydrogen production. The Copper-Chlorine (Cu-Cl) cycle is a prospective thermochemical cycle with a maximum temperature requirement of ~530°C. Thermalhydraulic calculations are presented for a double-pipe counter-flow heat exchanger with smooth pipe conditions and enhanced local heat transfer coefficients of 25%, 50% and 75% above smooth pipe cases. (author)

  20. Collaborative approach in developing a small supercritical water-cooled reactor

    International Nuclear Information System (INIS)

    A joint Research and Development (R and D) project between University of Saskatchewan and Atomic Energy of Canada (AECL) is being established to develop a concept of the small Canadian supercritical water-cooled reactor (SCWR) for power generation and process heat in remote areas. This project will be led by professors at the university and supported by technology experts from AECL. It integrates student training with a significant contribution to the reactor concept development. Students from various disciplines will combine results from physics, fuel, thermalhydraulic, control, material, and chemistry analyses to develop the core and fuel channel configurations and fuel design. This project would enhance the R and D expertise and capability of University of Saskatchewan and facilitate training of highly qualified persons (HQPs) for nuclear and non-nuclear industries at Saskatchewan and in Canada. (author)

  1. 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.

  2. Design of a containment vessel for a sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Reduction of plant construction cost is one of the most important issues for commercialization of fast reactors. From this point of view, an innovative containment vessel adopting steel plate reinforced concrete structure (SCCV) is developed for Japan Sodium-Cooled Fast Reactor (JSFR). Although SC structure is generally in practical use, performance after exposing high temperature is not investigated. An experimental study including loading and/or heating tests has been carried out to investigate the fundamental structural features, which would be provided to develop methodology to evaluate the feasibility of SCCV under the severe conditions. In this paper, the design feature, the design and evaluation conditions for SCCV of JSFR as well as the construction method are summarized. (author)

  3. 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.).

  4. 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)

  5. 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.

  6. 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.

  7. 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.

  8. Parameter estimation from dragon high temperature gas cooled reactor dynamic experiments

    International Nuclear Information System (INIS)

    Dynamic experiments were performed on the Dragon high temperature gas cooled reactor at full power, 20 MW. Both terminated ramp and pseudo-random chain code perturbations were applied to a control rod for two amplitudes of reactivity perturbation. Neutron flux and thermocouple signals were observed and recorded together with samples of the inherent noise with the reactor unperturbed. Frequency responses were deduced from the measurements and compared with previous sinusoidal frequency response measurements and theoretical predictions. A simplified model was constructed and optimized by least squares fitting of the equivalent response from the binary cross correlator to the model's output. These optimizations showed that a very simple feedback model is appropriate to Dragon and that a good estimate of the power/reactivity coefficient and temperature coefficient of reactivity may be made. (author)

  9. Performance of metal and oxide fuels during accidents in a large liquid metal cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cahalan, J.; Wigeland, R. (Argonne National Lab., IL (USA)); Friedel, G. (Internationale Atomreaktorbau GmbH (INTERATOM), Bergisch Gladbach (Germany, F.R.)); Kussmaul, G.; Royl, P. (Kernforschungszentrum Karlsruhe GmbH (Germany, F.R.)); Moreau, J. (CEA Centre d' Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance (France)); Perks, M. (UKAEA Risley Nuclear Power Development Establishment (UK)

    1990-01-01

    In a cooperative effort among European and US analysts, an assessment of the comparative safety performance of metal and oxide fuels during accidents in a large (3500 MWt), pool-type, liquid-metal-cooled reactor (LMR) was performed. The study focused on three accident initiators with failure to scram: the unprotected loss-of-flow (ULOF), the unprotected transient overpower (UTOP), and the unprotected loss-of-heat-sink (ULOHS). Emphasis was placed on identification of design features that provide passive, self-limiting responses to upset conditions, and quantification of relative safety margins. The analyses show that in ULOF and ULOHS sequences, metal-fueled LMRs with pool-type primary systems provide larger temperature margins to coolant boiling than oxide-fueled reactors of the same design. 3 refs., 4 figs.

  10. Numerical Simulation of Accident Scenario in High Temperature Gas Cooled (Pebble Bed) Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Peter, Geoffrey J. [Oregon Institute of Technology - Portland Center, Portland (United States)

    2012-03-15

    The accident scenario resulting from blockages due to the retention of dust in the coolant gas or from the rupture of one or more fuel particles used in the High Temperature Gas Cooled (Pebble Bed) Nuclear Reactors is considered in this paper. The next generation of Advanced High Temperature Reactors (AHTR), are considered for nuclear power production, and for high-temperature hydrogen production using nuclear reactors to reduce the carbon footprint. Blockages can cause LOCA variations in flow and heat transfer that may lead to hot spots within the bed that could compromise reactor safety. Therefore, it is important to know the void fraction distribution and the interstitial velocity field in the packed bed. The blockage for this numerical study simulated a region with significantly lower void than that in the rest of the bed. Finite difference technique solved the simplified continuity, momentum, and energy equations. Any meaningful outcome of the solution depended largely upon the validity of the boundary conditions. Among them, the inlet and outlet velocity profiles required special attention. Thus, a close approximation to these profiles obtained from an experimental set-up established the boundary conditions. This paper presents the development of the elliptic-partial equation for a bed of a bed of pebbles, and the solution procedure. The paper also discusses velocity and temperature profiles obtained from both numerical and experimental set-up, with and without effect of blockage. Based on the studies it is evident that knowledge of LOCA velocity and temperature distribution within the fuel element in a Pebble Bed Nuclear Reactor or AHTR is essential for reactor safety.

  11. Modern passive safety system for the advanced fast reactors with sodium cooling

    International Nuclear Information System (INIS)

    In fast reactors with sodium coolant it is possible to avoid serious damages of a core even at the heaviest scripts of development of accidents if to provide influence on reactance with the help of various type of Passive Safety System (PSS). With input of the PSS the reactor gets an additional negative feedback on the reactance, working at an output of the basic operational parameters (temperature, the coolant flow rate, power) for maximum permissible sizes. The scientific-technical and patent sources analysis has shown, that now it is already offered more than two hundred the various devices, capable to carry out functions of the fast reactors PSS. Comparison of various types of PSS is carried out under 9 generalized characteristics including: passivity, thresholdness, generation of efforts, inertia, multi-channels, stability to operational factors, refusal safety, simplicity and presentation, development conditions. For quantitative comparison of the device ''the perfection degree'' (K≤1) was defined as average size under 9 generalized characteristics. From the considered types of fast reactors PSS the most perfect now are fusible Lyophobic devices, basically meeting the requirements on all characteristics. Results of Lyophobic Passive Safety System development for the advanced fast reactors with sodium cooling are considered. Serviceability of the offered designs is proved experimentally at various operation temperatures on breadboard models sylphon devices and devices of type the sylphon-container with various lyophobic liquids: alloy Wuds (Tmt=80,0 deg C), an alloy lead-bismuth (Tmt=123,5 deg C), cadmium (Ttm=320,0 deg C), aluminium (Tym=660,0 deg C), developed Lyophobic Fusible Passive Safety System on excess of temperature are of interest for nuclear power installations of various type, first of all, as passive devices scram reactor and protection of the process equipment. (author)

  12. Preliminary Analysis of Scale Effect on Heat Removal Mechanism in Reactor Cavity Cooling System

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chan Soo; Park, Byung Ha; Bae, Y. Y. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The Reactor Cavity Cooling System (RCCS) maintains the concrete temperature below the design limit during normal operation. In an accident condition, the RCCS removes the residual heat, which amounts to 0.3 to 0.6 % of the full reactor power without additional electricity or water coolant supply. The criteria for the RCCS design are determined from the temperature limit of the reactor vessel material and concrete. Its governing heat transfer modes are the radiation across the reactor cavity and the buoyancydriven internal convection in the riser ducts installed around the reactor cavity. KAERI has conducted research on experimental verification of the RCCS coolability to make sure of the inherent safety of a VHTR. A difficulty in the full-scale test of a huge RCCS requires a reduced scale test, and the scaling is imperative in this case. This paper presents the GAMMA+ analysis about the scale effect on RCCS heat removal mechanism to check the validity of Bae et al.'s scaling analysis. In this paper, a GAMMA+ analysis was conducted to check the validity of the scaling law for the RCCS heat removal mechanism. The analytical results show that the scaling based on the Planck number is useful to extrapolate the reactor vessel temperature from the scale-down test results. Because the heat transfer regime of the scale down tests is different from that of the full scale condition, it requires a careful approach to analyze the convective heat transfer in the riser duct. In the present study, the comparison among the test results at ANL, KAERI, and UW will provide the information to develop and confirm the scaling criteria selected for an RCCS coolability demonstration.

  13. Conceptual design modifications of the cooling system of MNSR reactor to increase its maximum continuous operation time

    International Nuclear Information System (INIS)

    To increase the maximum daily operation time of Miniature Neutron Source Reactor (MNSR) reactor several conceptual thermal hydraulic design modifications have been investigated aiming at the improvement of reactor cooling conditions to limit the increase of average core temperature. For this purpose an integrated full-scale thermal hydraulic-neutronics model using the advanced code ATHLET has been developed, tested and verified. The selected design modifications rely upon introducing auxiliary cooling systems operating in four different modes to cool pool water or reactor water using heat exchanger located either inside or outside of reactor pool. The simulation results show that the increase of continuous reactor operation time varies between 1 and 8 additional operation hours. The optimal results are achieved for the second and the fourth options that use external heat exchanger. The second option enables the extending of continuous operation time up to 10 h and the fourth up to 15 h, both at nominal reactor power and under the assumption of initial excess reactivity corresponding to the fresh reactor core. The analysis included the evaluation of xenon poisoning effect on the increase of operation time. It has been shown that its remarkable effect starts after the first 3 operation hours and increases continuously after that. For the best cooling options, where the average core temperature is being fixed at certain value resulting in complete elimination of reactivity feedback of cooling temperature, xenon effect becomes the exclusive limiting effect during the later operation phase. The analysis discuss also general aspects of technical realization for the different cooling options in relation with the specific features of MNSR and the preliminary engineering safety measures and operational radiological protection that have to be taken. The performed analysis and the achieved results during this work would make valuable contribution for updating the Safety

  14. Factors affecting the performances of sprayed chromium carbide coatings for gas-cooled reactor heat exchangers

    International Nuclear Information System (INIS)

    The paper discusses some important factors to be considered for using sprayed coatings in gas-cooled reactor heat exchangers. These factors include (a) high-temperature gaseous corresion, (b) thermal stability of coatings, (c) metallurgical compatibility between the coating and substrate, and (d) effects of the coating on the mechanical properties of the substrate alloy. The coatings evaluated were Cr3C2--NiCr and Cr23C6--NiCr applied by either plasma-arc or detonation-gun process

  15. 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.

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

    International Nuclear Information System (INIS)

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

  17. 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.

  18. Specialists' meeting on heat exchanging components of gas-cooled reactors

    International Nuclear Information System (INIS)

    The objective of the Meeting sponsored by IAEA was to provide a forum for the exchange and discussion of technical information related to heat exchanging and heat conducting components for gas-cooled reactors. The technical part of the meeting covered eight subjects: Heat exchanging components for process heat applications, design and requirements, and research and development programs; Status of the design and construction of intermediate He/He exchangers; Design, construction and performance of steam generators; Metallic materials and design codes; Design and construction of valves and hot gas ducts; Description of component test facilities and test results; Manufacturing of heat exchanging components

  19. Thermal-hydraulics and safety concepts of supercritical water cooled reactors

    International Nuclear Information System (INIS)

    The paper summarizes the status of safety system development for supercritical water cooled reactors and of thermal-hydraulic codes needed to analyze them. While active safety systems are well understood today and expected to perform as required, the development of passive safety systems will still need further optimization. Depressurization transients have successfully been simulated with some codes by a pseudo-two-phase flow simulation of supercritical water. Open issues of thermal-hydraulic codes include modeling of deteriorated heat transfer in one-dimensional system codes and predictions of heat transfer during depressurization transients from supercritical to sub-critical conditions. (author)

  20. 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.

  1. High-temperature gas-cooled-reactor steam-methane reformer design

    International Nuclear Information System (INIS)

    The concept of the long distance transportation of process heat energy from a High Temperature Gas Cooled Reactor (HTGR) heat source, based on the steam reforming reaction, is currently being evaluated as an energy source/application for use early in the 21st century. The steam-methane reforming reaction is an endothermic reaction at temperatures approximately 7000C and higher, which produces hydrogen, carbon monoxide and carbon dioxide. The heat of the reaction products can then be released, after being pumped to industrial site users, in a methanation process producing superheated steam and methane which is then returned to the reactor plant site. In this application the steam reforming reaction temperatures are produced by the heat energy from the core of the HTGR through forced convection of the primary or secondary helium circuit to the catalytic chemical reactor (steam reformer). This paper summarizes the design of a helium heated steam reformer utilized in conjunction with a 1170 MW(t) intermediate loop, 8500C reactor outlet temperature, HTGR process heat plant concept. This paper also discusses various design considerations leading to the mechanical design features, the thermochemical performance, materials selection and the structural design analysis

  2. Design of the material performance test apparatus for high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Most materials can be easily corroded or ineffective in carbonaceous atmospheres at high temperatures in the reactor core of the high temperature gas-cooled reactor (HTGR). To solve the problem, a material performance test apparatus was built to provide reliable materials and technical support for relevant experiments of the HTGR. The apparatus uses a center high-purity graphite heater and surrounding thermal insulating layers made of carbon fiber felt to form a strong carbon reducing atmosphere inside the apparatus. Specially designed tungsten rhenium thermocouples which can endure high temperatures in carbonaceous atmospheres are used to control the temperature field. A typical experimental process was analyzed in the paper, which lasted 76 hours including seven stages. Experimental results showed the test apparatus could completely simulate the carbon reduction atmosphere and high temperature environment the same as that confronted in the real reactor and the performance of screened materials had been successfully tested and verified. Test temperature in the apparatus could be elevated up to 1600℃, which covered the whole temperature range of the normal operation and accident condition of HTGR and could fully meet the test requirements of materials used in the reactor. (authors)

  3. Concept on inherent safety in high-temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    A new safety concept in a high-temperature gas-cooled reactor (HTGR) was proposed to provide the most advanced nuclear reactor that exerts no harmful consequences on the people and the environment even if multiple failures in all safety systems occur. The proposed safety concept is that the consequence of the accidents is mitigated by the confinement of fission products employing not multiple physical barriers as in light water reactors, but only the cladding of fuel (i.e., the coating layers of the coated fuel particle). The progression of the events that lead to the loss or degradation of the confinement function of the coating layers (i.e., core heat up, oxidation of the coating layers, and explosion of carbon monoxide) is suppressed by only physical phenomena (i.e., the Doppler effect, thermal radiation and natural convection, formation of a protective oxide layer for coating layers of fuel, oxidation of carbon monoxide) that emerge deterministically as a cause of the events. The feasibility studies for severe events and related information revealed that the HTGR design based on this safety concept is technically feasible. This concept indicates the direction in which nuclear reactor research should be headed in terms of safety after the accident at the Fukushima Daiichi Nuclear Power Plant. (author)

  4. 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)

  5. 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.

  6. Control of Canadian once-through direct cycle supercritical water-cooled reactors

    International Nuclear Information System (INIS)

    Highlights: • Dynamic characteristics of Canadian SCWR are analyzed. • Hybrid feedforward and feedback control is adopted to deal with cross-coupling. • Gain scheduling control with smooth weight is applied to deal with nonlinearity. • It demonstrates through simulation that the control requirements are satisfied. - Abstract: Canadian supercritical water-cooled reactor (SCWR) can be modelled as a Multiple-input Multiple-output (MIMO) system. It has a high power-to-flow ratio, strong cross-coupling and high degree of nonlinearity in its dynamic characteristics. Among the outputs, the steam temperature is strongly affected by the reactor power and the most challenging to control. It is difficult to adopt a traditional control system design methodology to obtain a control system with satisfactory performance. In this paper, feedforward control is applied to reduce the effect on steam temperature from the reactor power. Single-input Single-output (SISO) feedback controllers are synthesized in the frequency domain. Using the feedforward controller, the steam temperature variation due to disturbances at the reactor power has been significantly suppressed. The control system can effectively maintain the overall system stability and regulate the plant around a specified operating condition. To deal with the nonlinearities, gain scheduling control strategy is adopted. Different sets of controllers combined by smooth weight functions are used for the plant at different load conditions. The proposed control strategies have been evaluated under various operating scenarios. Simulation results show that satisfactory performance can successfully achieved by the designed control system

  7. 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.

  8. Lead-cooled fast reactor (BREST) with an on-site fuel cycle

    International Nuclear Information System (INIS)

    Full text: Out of a great many of new power technologies, fission nuclear power is the only realistic way to stop the growth in extraction and combustion of fossil fuel. However, this is something to be achieved only through the nuclear power to have by the mid-21st century the capacity an order of magnitude as high as the current level. Nuclear power of such a scale will necessitate a new nuclear technology which is required to provide: transition to power with unlimited fuel resources; economically competitive nuclear power through reducing the cost of building and operating NPPs of a high inherent safety level with highly efficient utilization of fuel and generated heat; elimination of severe accidents with radioactive release which require evacuation of population, to be achieved primarily through combining inherent safety, passive protection features and impossible loss of lead coolant; an environmentally safe closed fuel cycle with in-pile combustion of minor actinides and radiation-equivalent disposal of radioactive waste; creation of nuclear proliferation barriers by way of eliminating uranium enrichment and plutonium separation facilities. These problems are solved in the BREST lead-cooled fast reactor. The initial stage plan is to build an NPP with a demonstration reactor and an on-site fuel cycle to verify designs, try out processes involving lead using as the coolant and study the behavior of the reactor and its systems and components in different modes, including resistance to anticipated operational occurrences and accidents simulated on the reactor. This will be followed by a rapid transition to a fast lead-cooled power reactor of 1200 MW(e) featuring two- circuit heat removal from the core to the turbine with supercritical steam parameters. The state of the activities to develop the NPP with lead-cooled fast reactors and an on-site fuel cycle is presented. A core with a moderate power rating has been considered. This will have a uranium

  9. Methods of Control-Rod Calibration in the Windscale Advanced Gas-Cooled Reactor

    International Nuclear Information System (INIS)

    Different techniques were used to calibrate control rods and to measure individual rod worths during the commissioning of the WAGR. These methods are described and the results are presented. The methods described are: (a ) Air poisoning - Changes in air pressure allow axial movement of rods at the critical condition. Thus rod movement can be related to pressure variation which is equivalent to a reactivity change. Also, rod slope measurements can be made for different rod insertions. (b) Rod slopes - The rods are withdrawn to make the reactor supercritical; then later they are inserted to make the reactor subcritical. From a measurement of the doubling and halving times the change in reactivity between the super- and subcritical states can be determined.- (c ) Absorber addition and withdrawal - The number of fixed localized absorbers is varied to give a method similar to the use of uniform air poisoning. (d) Pulsatron - This technique is used to give subcritical measurements of rod worth. (e) Rod run-in - The reactor is initially critical, and then the rods are run into the core. Analysis of the flux response relates reactivity to rod movement. (author)

  10. Conceptual design study of Pebble Bed Type High Temperature Gas-cooled Reactor with annular core structure

    International Nuclear Information System (INIS)

    This report presents the Conceptual Design Study of Pebble Bed Type High Temperature Gas-cooled Reactor with Annular Core Structure. From this study, it is made clear that the thermal power of the Pebble Bed Type Reactor can be increased to 500MW through introducing the annular core structure without losing the inherent safe characteristics (in the coolant depressurization accident, the fuel temperature does not exceed the temperature where the fuel defect begins.) This thermal power is two times higher than the inherent safe Pebble Bed Type High temperature Gas-cooled Reactor (MHTGR) designed in West Germany. From this result, it is foreseen that the ratio of the plant cost to the reactor power is reduced and the economy of the plant operation is improved. The reactor performances e.g. fuel burnup and fuel temperature are maintained in same level of the MHTGR. (author)

  11. Study on turbulence characteristics of free surface flow for cooling of fusion reactors, accelerator targets and reactor safety

    International Nuclear Information System (INIS)

    For the development of innovative fusion reactors, we examine the film flow along the first wall to simplify blanket and reduce the cost. A film flow is formed in primary cooling circuits of the light water reactors (LWR) when the loss of coolant accident (LOCA) occurs and a cold water is injected into the primary systems. In order to estimate the interfacial condensation rate at the developing region, it is required to have the knowledge about interfacial turbulent thermal diffusion of a thick film flow. Therefore, these systems have the same problem of heat transfer and transport inside the film flows. It is necessary to investigate the velocity and turbulence characteristics that have a close relation to the heat transfer and transport. Although there have been performed various studies on turbulence structure having free surface in a fully developed flow region, the turbulence properties of the film flows in a developing flow region has not been investigated sufficiently. Thus, we measure the velocity profiles and velocity fluctuations in a developing flow region using Laser Doppler Velocimeter (LDV). Then, experimental data are compared with analytical result that is obtained using the k-ε model of turbulence. (author)

  12. 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

  13. Observations of convective cooling in the tropical tropopause layer in AIRS data

    Directory of Open Access Journals (Sweden)

    H. Kim

    2004-11-01

    Full Text Available We investigate the impact of convection on the thermal structure of the Tropical Tropopause Layer (TTL. We use temperature profiles measured by the Atmospheric Infrared Sounder (AIRS onboard the Aqua satellite, and the time evolution of local convection determined by the National Centers for Environmental Protection/Aviation Weather Center (NCEP/AWS half-hourly infrared global geostationary composite. The observations demonstrate that the TTL is cooled by convection, in agreement with previous observations and model simulations. By using a global data set, we are able to investigate the variations in this convective cooling by season and region. The estimated cooling rate during active convection is 7.5~9 K/day. While we cannot unambiguously identify the cause of this cooling, our analysis suggests that radiative cooling is likely not an explanation.

  14. Observations of convective cooling in the tropical tropopause layer in AIRS data

    Science.gov (United States)

    Kim, H.; Dessler, A. E.

    2004-11-01

    We investigate the impact of convection on the thermal structure of the Tropical Tropopause Layer (TTL). We use temperature profiles measured by the Atmospheric Infrared Sounder (AIRS) onboard the Aqua satellite, and the time evolution of local convection determined by the National Centers for Environmental Protection/Aviation Weather Center (NCEP/AWS) half-hourly infrared global geostationary composite. The observations demonstrate that the TTL is cooled by convection, in agreement with previous observations and model simulations. By using a global data set, we are able to investigate the variations in this convective cooling by season and region. The estimated cooling rate during active convection is 7.5~9 K/day. While we cannot unambiguously identify the cause of this cooling, our analysis suggests that radiative cooling is likely not an explanation.

  15. 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.

  16. Experimental tests on the air cooling of the CLIC vertex detector

    CERN Document Server

    Duarte Ramos, Fernando; Nuiry, Francois-Xavier

    2016-01-01

    The strict requirements in terms of material budget for the inner region of the CLIC detector concept require the use of a dry gas for the cooling of the respective sensors. This, in conjunction with the compactness of the inner volumes, poses several challenges for the design of a cooling system that is able to fulfil the required detector specifications. This note summarizes the results obtained from experimental tests on the air cooling of the CLIC vertex detector as well as their comparison with the corresponding computational fluid dynamics simulations.

  17. Gas cooled fast reactor materials: compatibility and reaction kinetics of fuel/matrices couples

    International Nuclear Information System (INIS)

    Fourth Generation Gas cooled Fast Reactor concept implies a fast neutron spectrum and aims to lead to an iso-generation of minor actinides. Criteria have been defined for these fuels such as: high core filling factor, efficient fuel cooling, low operation temperature, i.e. 400-850 deg C, good fission product retention, burn-ups in the range of 5-8 atom%, Pu content in the range of 15-25%. Materials matching this demand are considered: mixed uranium - plutonium nitrides and carbides as fuels, whereas TiN, TiC, ZrN, ZrC, SiC are investigated as inert matrices. Thermo-chemical compatibility studies have been carried out, mostly for (U,Pu)N/SiC and (U,Pu)N/TiN couples. They have been associated to matching diffusional studies. For the first studies, accidental reactor conditions have been chosen (1600 deg C) so as to select a couple. Results are presented in terms of nature and quantity of resulting phases identified by XRD and SEM for thermodynamical equilibrium experiments. (authors)

  18. 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

  19. Comparative evaluation of pebble-bed and prismatic fueled high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kasten, P.R.; Bartine, D.E.

    1981-01-01

    A comparative evaluation has been performed of the HTGR and the Federal Republic of Germany's Pebble Bed Reactor (PBR) for potential commercial applications in the US. The evaluation considered two reactor sizes (1000 and 3000 MW(t)) and three process applications (steam cycle, direct cycle, and process heat, with outlet coolant temperatures of 750, 850, and 950/sup 0/C, respectively). The primary criterion for the comparison was the levelized (15-year) cost of producing electricity or process heat. Emphasis was placed on the cost impact of differences between the prismatic-type HTGR core, which requires periodic refuelings during reactor shutdowns, and the pebble bed PBR core, which is refueled continuously during reactor operations. Detailed studies of key technical issues using reference HTGR and PBR designs revealed that two cost components contributing to the levelized power costs are higher for the PBR: capital costs and operation and maintenance costs. A third cost component, associated with nonavailability penalties, tended to be higher for the PBR except for the process heat application, for which there is a large uncertainty in the HTGR nonavailability penalty at the 950/sup 0/C outlet coolant temperature. A fourth cost component, fuel cycle costs, is lower for the PBR, but not sufficiently lower to offset the capital cost component. Thus the HTGR appears to be slightly superior to the PBR in economic performance. Because of the advanced development of the HTGR concept, large HTGRs could also be commercialized in the US with lower R and D costs and shorter lead times than could large PBRs. It is recommended that the US gas-cooled thermal reactor program continue giving primary support to the HTGR, while also maintaining its cooperative PBR program with FRG.

  20. Comparative evaluation of pebble-bed and prismatic fueled high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    A comparative evaluation has been performed of the HTGR and the Federal Republic of Germany's Pebble Bed Reactor (PBR) for potential commercial applications in the US. The evaluation considered two reactor sizes [1000 and 3000 MW(t)] and three process applications (steam cycle, direct cycle, and process heat, with outlet coolant temperatures of 750, 850, and 9500C, respectively). The primary criterion for the comparison was the levelized (15-year) cost of producing electricity or process heat. Emphasis was placed on the cost impact of differences between the prismatic-type HTGR core, which requires periodic refuelings during reactor shutdowns, and the pebble bed PBR core, which is refueled continuously during reactor operations. Detailed studies of key technical issues using reference HTGR and PBR designs revealed that two cost components contributing to the levelized power costs are higher for the PBR: capital costs and operation and maintenance costs. A third cost component, associated with nonavailability penalties, tended to be higher for the PBR except for the process heat application, for which there is a large uncertainty in the HTGR nonavailability penalty at the 9500C outlet coolant temperature. A fourth cost component, fuel cycle costs, is lower for the PBR, but not sufficiently lower to offset the capital cost component. Thus the HTGR appears to be slightly superior to the PBR in economic performance. Because of the advanced development of the HTGR concept, large HTGRs could also be commercialized in the US with lower R and D costs and shorter lead times than could large PBRs. It is recommended that the US gas-cooled thermal reactor program continue giving primary support to the HTGR, while also maintaining its cooperative PBR program with FRG

  1. Modelling of subcooled boiling in ATHLET and application in water cooled research reactors

    International Nuclear Information System (INIS)

    A model is implemented to describe the thermodynamic nonequilibrium effects in subcooled boiling regime. The aim is to simulate void distribution and thermodynamic instability, which is practicularly pronounced in research reactors due to high power densities and low system pressures, and to include the influence of the steam formed in this boiling regime on the neutron balance. The model developed considers the competing effects of vaporization and condensation during subcooled boiling. It describes the rate of bubble generation on superheated surfaces and the subsequent condensation of steam in the subcooled liquid. The installed model is validated by postcalculations of two extensive series of experiments. The extended and verified program is used to simulate the Juelich research reactor FRJ-2. For this purpose, a full-scale simulation model of the entire plant is developed ensuring, in particular, a precise reproduction of the geometry and the arrangement of the annular fuel element cooling channels. The modelled reactor plant is first used to simulate normal reactor operation. The resulting steady-state temperature and pressure distributions assuming a thermal power of 23 MW show good agreement with real operating data. Safety investigations are conducted to examine plant behaviour under design-basis accident conditions. This includes failure of all three main coolent pumps with proper and delayed reactor scram. In both cases, the simulation shows that the fuel elements are not endangered in any phase of the transient, although in the event of a delayed scram initial signs of parallel channel instability due to steam formation in the central fuel element are to be observed which, however, only prevails for a short period of 30 ms. (orig./HP)

  2. Nuclear characteristics of a fissioning uranium plasma test reactor with light-water cooling

    Science.gov (United States)

    Whitmarsh, C. L., Jr.

    1973-01-01

    An analytical study was performed to determine a design configuration for a cavity test reactor. Test section criteria were that an average flux of 10 to the 15th power neutrons/sq cm/sec (E less than or equal to 0.12 eV) be supplied to a 61-cm-diameter spherical cavity at 200-atm pressure. Design objectives were to minimize required driver power, to use existing fuel-element technology, and to obtain fuel-element life of 10 to 100 full-power hours. Parameter calculations were made on moderator region size and material, driver fuel arrangement, control system, and structure in order to determine a feasible configuration. Although not optimized, a configuration was selected which would meet design criteria. The driver fuel region was a cylindrical annular region, one element thick, of 33 MTR-type H2O-cooled elements (Al-U fuel plate configuration), each 101 cm long. The region between the spherical test cavity and the cylindrical driver fuel region was Be (10 vol. % H2O coolant) with a midplane dimension of 8 cm. Exterior to the driver fuel, the 25-cm-thick cylindrical and axial reflectors were also Be with 10 vol. % H2O coolant. The entire reactor was contained in a 10-cm-thick steel pressure vessel, and the 200-atm cavity pressure was equalized throughout the driver reactor. Fuel-element life was 50 hr at the required driver power of 200 MW. Reactor control would be achieved with rotating poison drums located in the cylindrical reflector region. A control range of about 18 percent delta k/k was required for reactor operation.

  3. STOCHASTIC ANALYSIS OF AN AIR CONDITION COOLING SYSTEM MODEL

    Directory of Open Access Journals (Sweden)

    Punam Phartyal

    2011-01-01

    Full Text Available The present study deals with the stochastic analysis of a real existing industrial systemmodel of a central air-condition (AC system. The system consists of three different subsystemsnamely- Air Blower, Compressor, water pump. All these subsystems are arranged in seriesnetwork. Transition probabilities as well as the recurrence relations for various reliability andcost effective measures are developed. Failure time distributions of all the subsystems are takenas exponential whereas repair time distributions are general. By using regenerative pointtechnique we have obtained various measures of system effectiveness such as –Reliability,MTSF, Availability, Busy period of repairman and Net expected profit. The results are alsodrawn in a particular case when repair time distributions are assumed as exponentials.

  4. 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

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

    International Nuclear Information System (INIS)

    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 equipment. 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 homogeneously 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 occurred 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, expansion load, and combination load and the stress fraction have reached 95,5%. Therefore a pipeline modification (re-routing) 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 30, also a

  6. 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)

  7. 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.

  8. Control approach to the load frequency regulation of a Generation IV Lead-cooled Fast Reactor

    International Nuclear Information System (INIS)

    Highlights: • Dedicated control strategy for adjusting the electrical power according to the grid requirements. • Decoupling of the Balance of Plant from the reactor primary circuit thanks to effective feedback regulators. • Primary frequency regulation and islanding simulations assessed with an object-oriented model. - Abstract: One of the most pressing issues in the study of the power generation and distribution is the characterization of the grid behavior, whether a relevant fraction of the connected power plants relies on Renewable Energy Sources. Indeed, because of the discontinuous power supply and the limited presence of energy accumulators, concerning power imbalances may take place on the grid. The power plants ensuring high reliability performance should be ready to feed the loads when the Renewable Energy Sources are not available. In order to ensure the grid stability and the sustainability of nuclear energy, the possibility of operating Generation-IV nuclear reactors in a flexible way should be considered, i.e., the Nuclear Power Plants should adjust the mechanical power produced so as to comply with the sudden grid frequency variations. In the present work, this opportunity is assessed for the Lead-cooled Fast Reactors, adopting the Advanced Lead Fast Reactor European Demonstrator (ALFRED) as a representative of Lead-cooled Fast Reactor technology. For this reactor concept, because of the large thermal inertia that characterizes the system, the adoption of the “reactor-follows-turbine” scheme (currently employed in the Pressurized Water Reactors) is not feasible. An alternative solution is proposed, i.e., the set-point for the thermal power produced in the core is kept constant at the nominal value (or slowly variable), and the set-point for the mechanical power available to the alternator is adjusted according to the load demands. In order to assess the performance of the developed control scheme, two case studies are simulated. In the

  9. Remediation of a large contaminated reactor cooling reservoir: Resolving and environmental/regulatory paradox

    Energy Technology Data Exchange (ETDEWEB)

    Bowers, J.A.: Gladden, J.B.; Hickey, H.M.; Jones, M.P.; Mackey, H.E.; Mayer, J.J. [Westinghouse Savannah River Co., Aiken, SC (United States); Doswell, A. [USDOE, Washington, DC (United States)

    1994-05-01

    This paper presents a case study of a former reactor cooling water reservoir, PAR Pond, located Savannah River Site. PAR Pond, a 2640 acre, man-made reservoir was built in 1958 and until 1988, received cooling water from two DOE nuclear production reactors, P and R. The lake sediments were contaminated with low levels of radiocesium (CS-137) and transuranics in the late 1950s and early 1960s because of leaking fuel elements. Elevated levels of mercury accumulated in the sediments from pumping water from the Savannah River to maintain a full pool. PAR Ponds` stability, size, and nutrient content made a significant, unique, and highly studied ecological resource for fish and wildlife populations until it was partially drained in 1991 due to a depression in the downslope of the earthen dam. The drawdown, created 1340 acres of exposed, radioactively contaminated sediments along 33 miles of shoreline. This led US EPA to declare PAR Pond as a CERCLA operable unit subject to remediation. The drawdown also raised concerns for the populations of aquatic plants, fish, alligators, and endangered species and increased the potential for off-site migration of contaminated wildlife from contact with the exposed sediments. Applicable regulations, such as NEPA and CERCLA, require wetland loss evaluations, human health and ecological risk assessments, and remediation feasibility studies. DOE is committed to spending several million dollars to repair the dam for safety reasons, even though the lake will probably not be used for cooling purposes. At the same time, DOE must make decisions whether to refill and expend additional public funds to maintain a full pool to reduce the risks defined under CERCLA or spend hundreds of millions in remediation costs to reduce the risks of the exposed sediments.

  10. Preliminary evaluation of steam generator tube rupture (SGTR) accident in lead cooled reactor

    International Nuclear Information System (INIS)

    In this paper some contributions are provided to the development of a European Lead-cooled System, known as the ELSY project (within EU-6 Framework Project); that will constitute a possible reference system for a large lead-cooled reactor of GEN IV. Steam generator (SG) tubing of this system type might be subject to a variety of degradation processes, such as cracking, wall thinning and potential leakage or rupture, eventually leading to the failure of one or more SG tubes that constitute a steam generator tube rupture (SGTR) accident with possible consequences for the safety of the primary systems. It is therefore of interest for the designer to know how the SG itself, as well as the vessel and internals structures, behave under impulsive loading conditions (in form of a rapid and strong increase of pressure) that can arise as consequences of the interaction between the primary and secondary coolants (lead-water interaction). The analysed initiator event, as already mentioned, is a large break (up to a double ended guillotine break) of one (or more) SG cooling tubes that may become severe enough to determine dangerous effects on the interested structures. In order to better simulate and perform the mentioned postulated SGTR accident sequence analyses, an appropriate numerical model with the available computing resources (FEM codes) was set up at the DIMNP of Pisa University. That model was used to evaluate the effects of the propagation of the blast pressure waves inside the SG structures, taking into account also the sloshing phenomenon that could be induced by the lead primary coolant motions. Therefore the SGTR effects study may be considered as a transient and non linear problem the solution of which provides the 'time histories' of hydrodynamic pressures and stresses on the reactor pressure vessel and internals walls. (author)

  11. 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.

  12. 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

  13. 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

  14. Technical meeting on 'Primary coolant pipe rupture event in liquid metal cooled fast reactors'. Working material

    International Nuclear Information System (INIS)

    In Liquid Metal cooled Fast Reactors (LMFR) or in accelerator driven sub-critical systems (ADS) with LMFR like sub-critical cores, the primary coolant pipes (PCP) connect the primary coolant pumps to the grid plate. A rupture in one of these pipes could cause significant loss of coolant flow to the core with severe consequences. In loop type reactors, all primary pipelines are provided with double envelopes and inter-space coolant leak monitoring systems that permit leak detection before break. Thus, the PCP rupture event can be placed in the beyond design basis event (BDBE) category. Such an arrangement is difficult to incorporate for pool type reactors, and hence it could be argued that the PCP rupture event needs to be analysed in detail as a design basis event (DBE, category 4 event). The primary coolant pipes are made of ductile austenitic stainless steel material and operate at temperatures of the cold pool and at comparatively low pressures. For such low stressed piping with negligible creep and embrittlement effects, it is of interest to discuss under what design provisions, for pool type reactors, the guillotine rupture of PCP could be placed in the BDBE category. The topical Technical Meeting (TM) on 'Primary Coolant Pipe Rupture Event in Liquid Metal Cooled Reactors' was called to enable the specialists to present the philosophy and analyses applied on this topic in the various Member States for different LMFRs. The scope of the Technical Meeting was to provide a global forum for information exchange on the philosophy applied in the various participating Member States and the analyses performed for different LMFRs with regard to the primary coolant pipe rupture event. More specifically, the objectives of the Technical Meeting were to review the safety philosophy for the PCP rupture event in pool type LMFR, to assess the structural reliability of the PCP and the probability of rupture under different conditions (with/without in-service inspection), to

  15. 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

  16. Separate-effects experiments on the hydrodynamics of air ingress phenomena for the very high temperature reactor

    International Nuclear Information System (INIS)

    The present study performs scaled separate-effects experiments to investigate the hydrodynamics in the air-ingress phenomena following a Depressurized Condition Cooldown in the Very High Temperature Gas-Cooled Reactor. First, a scoping experiment using water and brine is performed. The volumetric exchange rate is measured using a hydrometer, and flow visualizations are performed. Next, Helium-air experiments are performed to obtain three-dimensional oxygen concentration transient data using an oxygen analyzer. It is found that there exists a critical density difference ratio, before which the ingress rate increases linearly with time and after which the ingress rate slows down significantly. In both the water-brine and Helium-air experiments, this critical ratio is found to be approximately 0.7. (author)

  17. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    Burrill, K.A

    2001-08-01

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

  19. 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.

  20. Research and development program of hydrogen production system with high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Japan Atomic Energy Research Institute (JAERI) has been developing a hydrogen production system with a high temperature gas-cooled reactor (HTGR). While the HTGR hydrogen production system has the following advantages compared with a fossil-fired hydrogen production system; low operation cost (economical fuel cost), low CO2 emission and saving of fossil fuel by use of nuclear heat, it requires some items to be solved as follows; cost reduction of facility such as a reactor, coolant circulation system and so on, development of control and safety technologies. As for the control and safety technologies, JAERI plans demonstration test with hydrogen production system by steam reforming of methane coupling to 30 Wt HTGR, named high temperature engineering test reactor (HTTR). Prior to the demonstration test, a 1/30-scale out-of-pile test facility is in construction for safety review and detailed design of the HTTR hydrogen production system. Also, design study will start for reduction of facility cost. Moreover, basic study on hydrogen production process without CO2 emission is in progress by thermochemical water splitting. (orig.)